u.s. NUCLEARREGULATORY COMMISSION

1979

Annual Report

The President
The White House
Washington, D.C. 20500

Dear Mr. President:
Enclosed is the fifth Annual Report of the United States
Nuclear Regulatory Commission for your trarnmittal to theCongress, as required by Section 307(c) of the Energy Reorg~­nization Act of 1974. This report covers the major activities of
the NRC from October 1, 1978 through September 30, 1979 andbriefly describes some additional actiorn through 1979 into 1980.

Respectfully, 'John F. Ahearne
Chairman

1979
Annual Reportu.s. NUCLEARREGULATORY COMMISSION

For sale by the Superintendllnt of Documents, u.s. Government Printing 0 lftce
Washington. D.C. 20402

iii

Table of Contents
Chapter I-OVERVIEWIntroduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
HIGHLIGHTS AND UPDATES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Accident at Three Mile Island . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
NRC Organization and Management . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Reactor Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Emergency Preparedness ....................... . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Analysis of Operating Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Inspection and Enforcement. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Radiation Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Research. . . . . . . . . . . . .. . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .
Waste Management . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Transportation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Domestic Safeguards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Exports and International Safeguards. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
THE NRC COMMITMENT.....................................................

1
2
2
3
4
5
5
6
6
6
7
8
8
9
9

Chapter 2-ACCIDENT AT THREE MILE ISLAND
WHAT HAPPENED. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Wednesday-March 28 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Thursday-March 29 ...................................... , , , , ... , . . . . . . . ..
Friday-March 30 .. , ............................................. , . . . . . . . ..
Saturday-March 31 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Sunday-April 1 ...........................................................
RADIOLOGICAL CONSEQUENCES TO PERSONS AND THE ENVIRONMENT. . . . ..
TMI RECOVERY OPERATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
BULLETINS AND ORDERS TASK FORCE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Bulletins ..................................................................
Orders on Babcock and Wilcox Plants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Specific Plant and Generic Studies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Follow-on and Interfacing Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Actions Directed by NRC Orders. . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Longer-term Actions Required by Orders. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
TMI-2 LESSONS LEARNED TASK FORCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Two Phases of Work . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Short-term Recommendations .............................. , . , . . . . . . . . . . . . . ..
Implementing Short-term Recommendations ............ , .. , .... , . . . . . . . . . . . . . ..
Long-term Recommendations . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Inspection and Enforcement Lessons. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
THE PRESIDENT'S COMMISSION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Findings and Judgments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Supplemental Views ................................................... , . , ..
Recommendations and Responses , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
PRESIDENT'S COMMISSION RECOMMENDATIONS ON THE NRC . . . . . . . . . . . . . . ..

11
12
17
17
18
19
20
22
24
25
25
26
26
27
27
27
27
28
30
30
32
35
35
39
42
43

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PRESIDENT'S COMMISSION RECOMMENDATIONS ON THE UTILITY ANDITS SUPPLIERS .............................................................
PRESIDENT'S COMMISSION RECOMMENDATIONS ON THE TRAINING
OF OPERATING PERSONNEL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
PRESIDENT'S COMMISSION RECOMMENDATIONS BASED ON ITSTECHNICAL ASSESSMENT..................................................
PRESIDENT'S COMMISSION RECOMMENDATIONS ON WORKER AND
PUBLIC HEALTH AND SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
PRESIDENT'S COMMISSION RECOMMENDATIONS ON EMERGENCY PLANNING
AND RESPONSE ............................................................
PRESIDENT'S COMMISSION RECOMMENDATIONS ON THE PUBLIC'S RIGHTTO INFORMATION .........................................................
The President's Response ................................................... "

48
50
51
53
54
56
60

Chapter 3-REACTOR REGULATION
Status of Nuclear Power Generation .......................................... .
Assistance From Other Agencies .............................................. .
THE LICENSING PROCESS ................................................... .
Licensing Reactor Operators .................................................. .
UNRESOL VED SAFETY ISSUES ............................................... .
Evaluation Process ......................................................... .
Identification of New Issues ................................................. .
Progress Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . ...... .
Water Hammer ........................................................... .
Asymmetric Blowdown Loads on the Reactor Coolant System ..................... .
PWR Steam Generator Tube Integrity ........................................ .
BWR Mark I and Mark II Pressure Suppression Containments ..................... .
Anticipated Transients Without Scram ........................................ .
BWR Nozzle Cracking ..................................................... .
Reactor Vessel Material Toughness ........................................... .
Fracture Toughness and Potential for Lamellar Tearing of PWR Steam Generator
and Reactor Coolant Pump Supports ........................................ .
Systems Interactions in Nuclear Power Plants ................................... .
Environmental Qualification of Safety-Related Electrical Equipment .. " .......... .
Reactor Vessel Pressure Transient ............................................ .
Residual Heat Removal Shutdown Requirements ................................ .
Control of Heavy Loads Near Spent Fuel ...................................... .
Seismic Design Criteria ..................................................... .
Pipe Cracks at Boiling Water Reactors ........................................ .
Containment Emergency Sump Reliability ..................................... .
Station Blackout .............................................. ; ........... .
OTHER TECHNICAL ISSUES .................................................. .
Design Errors in Control BUilding ............................................ .
PWR Feedwater Line Cracks ................................................ .
Mechanical Operability of Containment Purge Valves ........................... .
Environmental Qualification of Safety-Related Electrical Equipment at
Operating Plants ........ , . , ......................... , ................... .
Pipe Support Base Plate Problem , .. , ... , ..................................... .
Nonconformance of Actual Installation of Safety-Related Piping Systems to
Design Documents ....................................................... .
Shutdown and Seismic Reanalysis of Five Operating Reactors ..................... .
Petition for the Seismic Reanalysis of Operating Reactors ......................... .
Humboldt Bay .................................................... , .. , , ... .
Possible Faulting Near Reactor ....................... , ....................... .
Fire Protection ............................................................ .
Control Rod Guide Tube Integrity ................................. , ........ .
PROGRESS IN STANDARDIZATION ........................................... .
ADV ANCED NUCLEAR POWER PLANTS ....................................... .
Clinch River Breeder Reactor .......................................... , . . . . ..

63
63
64
65
65
66

66
67
67
68
69
71
73
73

75
76
77

79
80
80
81

82
82
85

86
86
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89
89

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91

93
93
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94

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98
98

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Fast Flux Test Facility ................................................ , . . . . .. 98
Gas~Cooled Reactors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
99
Floating Nuclear Power Plants. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 99
PROTECTING THE ENVIRONMENT ........................................... 100
Health Effects of the Coal and Nuclear Fuel Cycles. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 100
Assessment of Radiological Consequences of Radionuclide Releases . . . . . . . . . . . . . . . . .. 100
Control of Effluents. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 101
Site-Related Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 101
INTERAGENCY COORDINATION .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 104
Compliance with Regulations of Other Agencies. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 104
Interagency Topical Studies. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 105
COOPERATION WITH STATES ................................................ 107
State Participation in NRC's Environmental Impact Statements. . . . . . . . . . . . . . . . . . .. 107
IMPROVING THE LICENSING PROCESS ........................................ 107
Generic Rulemaking to Improve Licensing. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 107
Evaluation of Alternative Sites. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 108
Siting Policy Task Force . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 108
Early Site Reviews. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 110
Environmental Standard Review Plans .......................................... 110
Environmental Impact Statements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 110
Social and Economic Issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 110
Information Base for Licensing Decisions ..................................... " 112
Quality Assurance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 113
Systematic Evaluation of Operating Reactors. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 114
ANTITRUST ACTIVITIES ...................................................... 115
INDEMNITY AND FINANCIAL PROTECTION . . .. . . . . . . . . . . . . . . . . . . .. . . . . . . . . . .. 116
Increase in Levels of Required Financial Protection. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 116
Financial Protection for Three Mile Island Units 1 and 2 . . . . . . . . . . . . . . . . . . . . . . . . . .. 116
Indemnification of Storage of Spent Fuel at Distant Reactor Locations . . . . . . . . . . . . . .. 117
Claims Handling Procedures Following TMI . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 117
Determination of an Extraordinary Nuclear Occurrence. . . . . . . . . . . . . . . . . . . . . . . . . .. 118
Indemnity Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 118
Insurance Premium Refund . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 118
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS . . . . . . . . . . . . . . . . . . . . . . . . .. 118

Chapter 4-MATERIALS REGULATIONFUEL CYCLE ACTIVITIES. . . . . . . . . . . .. . . . . . . . . . . . . . . . . . .. . .. .. . . . . . .. .. .. .. . .. 121
URANIUM FUEL CYCLE SURVEY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 122
SPENT FUEL STORAGE ACTIONS. . . . . . . . . . . . . . . . . . . . .. . .. . . . .. . .. . . .. .. .. . . . .. 123
Final Environmental Statement. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 123
Movements Between Reactors .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 123
Away-from-reactor Storage. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 123
ADV ANCED FUEL ACTIVITIES . . . . . . . . . . . . . . . . . . . .. . . . . .. . . .. . . .. . .. . . .. .. . ... 124
Review of Plutonium Plants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 124
OTHER FUEL CYCLE ACTIVITIES .. . . .. . . . . . . . . . . .. . . . . . . . .. .. . . .. .. .. .. . . . . .. 125
Evaluating Sites for Radioactivity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .125
West Valley, N. Y., Facility .................................................. 125
TRANSPORTATION OF RADIOACTIVE MATERIALS ............................. 126
Low-Level Waste Shipments ................................................. 126
Irradiated Fuel Packaging . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 127
Safety of Transportation Workers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 128
Transportation in Urban Areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 128
Emergency Response Planning . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 129
Packaging Standards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 129
International Standards. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 129
RADIOACTIVE MATERIAL LICENSING . . . . . .. . .. . .. . . . . . . . . . . . . . . . .. .. . . .. . . .. 130
Nuclear Medicine .......................................................... 1'31

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Chapter 5-DOMESTIC SAFEGUARDS
Scope of NRC Programs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 133
Assessment of Safeguards Adequacy Policy. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 133
Fuel Cycle Facilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 134
Transportation Activities .................................................... 135
Reactor Safeguards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 138
Contingency Planning. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 139
SAFEGUARDS EVENTS-FISCAL YEAR 1979 .................................... 139
SAFEGUARDS REGULATORY ACTIVITIES ...................................... 140
SAFEGUARDS RESEARCH AND TECHNICAL ASSISTANCE ....................... 142
SAFEGUARDS MANAGEMENT ................................................. 144

Chapter 6 -WASTE MANAGEMENT
Overview of 1979 Activity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 145
HIGH-LEVEL WASTE MANAGEMENT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 147
Regulatory Development .................................................... 147
Licensing ................................................................. 148
Assessment of DOE High-Level Waste Management Program ...................... 148
LOW-LEVEL WASTE MANAGEMENT .......................................... 149
Regulatory Development ............................. . . . . . . . . . . . . . . . . . . . . . .. 149
Licensing ................................................................. 149
Technical Assistance to Agreement States. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 151
URANIUM RECOVERY MANAGEMENT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 151
Regulatory Development ...... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 151
Licensing ................................................................. 151
Technical Assistance to Agreement States ....................................... 151
NRC Assessment of DOE Remedial Action Plans . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 152

Chapter 7-INSPECTION AND ENFORCEMENT
THE INSPECTION PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 153
Reporting Defects and Noncompliance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . .. 155
Types of Inspections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 155
Licensee, Contractor and Vendor Inspection Program . . . . . . . . . . . . . . . . . . . . . . . . . . .. 155
Performance Appraisal Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 157
Independent Measurement/Verification Program ................................ 157
Government-Industry Efforts. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 157
Inspection Activities Resulting from TMI . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 157
Impact of TMI on Inspection Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 158
Resident Inspector Program .................................................. 159
Bulletins, Circulars and Information Notices .................................... 159
Other Reactive Effort. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 166
ENFORCEMENT ACTIVITIES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .................. 166
Enforcement Improvements, ................................................... 168
NRC Operations Center ..................................................... 168
INVESTIGATIONS ............................................................ 168
Wolf Creek Generating Station. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 169
Marble Hill Nuclear Generating Station. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 170
McGuire Nuclear Plant. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 170
Midland Nuclear Plant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 170
Surry Nuclear Power Station. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 171
Abnormal Occurrences-Fiscal Year 1979 ...................................... 171

vii

Chapter 8-STATE PROGRAMSST ATE AGREEMENTS PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 173
Review of State Regulatory Programs .......................................... 173
Adequacy and Compatibility Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 173
NRC Technical Assistance ................................................... 174
Training Offered by NRC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 174
Annual Meeting ................................................... '. . . . . . . .. 175
Agreement States and Mill Tailings . . . . . . . . . . . . . . . . . . . . . . .. . ................ ". 175
Agreement State Abnormal Occurrences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 175
EMERGENCY PREPAREDNESS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 176
Emergency Response Planning ....................... , ........................ 176
Planning Guidance to States. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 177
Training Program for States. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 177
Field Assistance Program .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 178
TMI Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 178
GAO Report ............................................................... 178
Other Emergency Response Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 178
LIAISON AND COOPERATIVE ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 179
Transportation Surveillance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 179
Memorandums of Agreement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 179
State Liaison Officers Program .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 179
National/State Organizations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 180
Conference of Radiation Control Program Directors .............................. 180
Intergovernmental Personnel Assignments ...................................... 180
Workshops ................................................................ 181
Three Mile Island Waste Shipments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 181

Chapter 9-INTERNATIONAL ACTIVITIESINFORMATION EXCHANGES .................................................. 183
Bilateral Arrangements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 183
New Research Agreements ................................................... 184
Increase in Foreign Visitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 185
IAEA Nuclear Safety Program ................................................ 185
Technical Assistance through IAEA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 186
Cooperation with the OECD . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 186
EXPORT/IMPORT MATTERS AND NONPROLIFERATION . . . . . . . . . . . . . . . . . . . . . . .. 186
Retransfers for Reprocessing. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 187
Technology Transfers ....................................................... 187
Agreements for Cooperations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 187
Nuclear Fuel Cycle Evaluations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 187
Pacific Basin Fuel Storage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 188
NRC Views on Nonproliferation Role .......................................... 188
EXPORT LICENSING ACTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 189
Philippines Reactor Project ................................................... 189
Tarapur (India) Proceeding .................................................. 193
ENVIRONMENTAL EFFECTS OF EXPORTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 194
Unified Interagency Procedures ............................................... 194
Research and Development Reactor Exports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 194
INTERNATIONAL SAFEGUARDS ............................................... 194
US/IAEA Safeguards Agreement .............................................. 194
Export Licensing Information Needs. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 195
Support of International Safeguards ........................................... 195
Other Activities ............................................................ 195

viii
Chapter IO-STANDARDS DEVELOPMENT
CONCERNS OF HIGH PRIORITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 197
POWER REACTOR STANDARDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 200
Surveillance and Inservice Inspection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 200
Accident Analysis ........................................................... 201
Reactor Containment ....................................................... 201
System and Component Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ......... 202
Quality Assurance . . . . . . . . . . . . . . . . . . . . . . . . . .. .............................. 202
Qualification Testing ................................... . . . . . . . . . . . . . . . . . . .. 203
Electric Systems and Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 203
Reporting Defects and Noncompliance ......................................... 204
Protection Against Fire . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 204
Medical Certification ....................................................... 204
FUEL CYCLE PLANT STANDARDS ............................................. 204
Decommissioning .......................................................... 204
Spent Fuel Storage. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 205
Uranium Milling and Processing .............................................. 205
Nuclear Criticality Safety. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 205
Plant Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ............ 205
Waste Management . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 206
General Site Suitability Criteria. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 206
SITING STANDARDS .......................................................... 206
Site Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 206
Site Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 207
ENVIRONMENT AL PROTECTION ........................... ................. 207
Interagency Coordination ................................................... 208
RADIOLOGICAL HEALTH STANDARDS ........................................ 208
Low-Level Radiation Effects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 208
Nuclear Medicine. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . ........... 209
OCCUPATIONAL HEALTH STANDARDS ........................................ 210
Respiratory Protection ...................................................... 210
Testing for Personnel Dosimetry. . . . . . . . . . . . . . . . . . . . .. . ...................... 210
Personnel Monitoring Reports ................................................ 210
Worker Exposure to Neutrons ................................................ 211
Calibration of Air Sampling Instruments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ... 211
Transient Worker Radiation Protection ........................................ 211
Medical Institutions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 211
Bioassays . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ....... 212
Health Protection at Uranium Mills ............................................ 212
Effecting Occupational ALARA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 212
Surveys During U-235 Processing and Fuel Fabrication. . . . . . . . . . . . . . .. ........... 212
Radiation Safety Training at L WRs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 213
Ga"mma Irradiators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 213
Ind.u~trial ~~dio~raphy Saf~~y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 213
PetItIOn on RadIatIon Area ................................................. 213
SAFEGUARDS STANDARDS .................................................... 213
Physical Protection .............. .......................................... 213
Material Control and Accounting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 215
RADIOISOTOPES IN INDUSTRY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 215
Products Containing Radioactive Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 215
Licensing Matters ................................................ ......... 216
NATIONAL STANDARDS PROGRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 216
IAEA REACTOR SAFETY STANDARDS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 216

Chapter II-REGULATORY RESEARCH
WATER REACTOR SAFETY RESEARCH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
SYSTEMS ENGINEERING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Integral Systems Tests ............................. , . . . . . . . . . . . . . . . . . . . . . . . ..
Separate-Effects Experiments .................................................

219
219
220
222

·

IX

FUEL BEHAVIOR ............................................................. 224
Cladding Experiments ...................................................... 224
In-Reactor Testing. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 225
Fuel Meltdown/Fission Product Release and Transport ............................ 226
Fuel Behavior Codes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 226
COMPUTER CODE DEVELOPMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 227
Code Development . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 227
Code Assessment ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 228
METALLURGY AND MATERIALS ............................................. 228
Fracture Mechanics . . . . . . . . . . . . . .. .................. .. ................... 229
Steam Generator Tube Integrity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 230
Radiation Embrittlement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 230
Flaw Detection .............................................. . ............ 230
RESEARCH SUPPORT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 230
Operational Safety Research ................................................. 231
Program Support . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 233
Technical Support . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 233
ADV ANCED REACTOR RESEARCH ............. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 233
HIGH TEMPERATURE GAS-COOLED REACTORS.............................. 233
LIQUID MET AL~COOLED FAST BREEDER REACTORS . . . . . . . . . . . . . . . . . . . . . . . . .. 234
Analysis Program ......................................... ". . . . . . . . . .. ... .. 234
Safety Test Facilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 234
Materials Interactions ..................................................... " 234
Aerosol Release and Transport . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 234
Systems Integrity. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 235
GENERAL REACTOR SAFETY RESEARCH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 235
SITE SAFETY RESEARCH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .................... 235
Geology and Seismolo~ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 235
236
Meteorology and Hydrology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
MECHANICAL ENGINEERING RESEARCH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 236
STRUCTURAL ENGINEERING RESEARCH. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 237
FUEL CYCLE, ENVIRONMENTAL AND WASTE MANAGEMENT RESEARCH ....... 238
FUEL CYCLE RESEARCH ............ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 238
ENVIRONMENT AL RESEARCH ....................... . . . . . . . . . . . . . . . . . . . . . . . .. 238
Radiation Dosimetry and Health Effects. . . . . . . . . . . . . . . . . . . . . . . . . . . ..... . . . .. 238
Ecological Impact Studies. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ... . 238
Environmental Transport and Effluent Monitoring. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 239
Socioeconomic Impacts and Regional Siting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 23B
WASTE MANAGEMENT RESEARCH. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 239
High-Level Radioactive Waste Research. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 239
Low-Level Radioactive Waste Research . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 240
Uranium Milling Research Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 240
RISK ASSESSMENT RESEARCH ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 240
Lewis Group Review of WASH-I400 .......................................... 240
Reactor Accident Consequence Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 241
Reactor Risk Assessment and Licensing Support . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 241
Fuel Cycle Risk Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . .. 241
Methodolo~ry Development .................................................. 241
RESEARCH TO IMPROVE REACTOR SAFETY . . . . . . . . . . . . . . . . .. ................ 242
243
Improved In-Plant Accident Response. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Alternate Containment Concepts. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 243
Alternate Decay Heat Removal Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . .. 243
Improved Methodology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 243

Chapter 12-INFORMING AND INVOLVING THE PUBLIC
Making Documents Available ................................................
INVOL VING THE PUBLIC. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..
Informal Participation ............................. . . . . . . . . . . . . . . . . . . . . . . . ..
Formal Public Participation ................................................ "ENHANCING INTERNAL COMMUNICATION . . . . . . . . . . . . .. ... ... ... ........
CONGRESSIONAL OVERSIGHT .............................................. "Reports to Congress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..

245
247
247
249
250
250
253

x
Chapter 13-PROCEEDINGS AND LITIGATIONATOMIC SAFETY AND LICENSING BOARDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 255
ATOMIC SAFETY AND LICENSING APPEAL BOARDS ............................ 258
COMMISSION DECISIONS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 260
Final Fuel Cycle Rule Promulgated . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 260
Three Mile Island Unit 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 260
Offshore Power Systems ..................................................... 261
Waste Confidence Rulemaking ............................................... 261
Bailly Short Pilings Petitions Denied ........................................... 261
JUDICIAL REVIEW ........................................................... 262
Significant Cases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 262
Concluded Cases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 264
Opened Cases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 267
Pending Cases ............................................................. 268

Chapter 14-ADMINISTRATION AND MANAGEMENTPERSONNEL AND ORGANIZATION . . . . . • . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 273
Commission and Office Director Appointments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 273
Organizational Changes ..................................................... 273
MANAGEMENT -EMPLOYEE RELATIONS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 275
Civil Service Reform Act ..................................................... 275
Union Activity. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 275
Equal Employment Opportunity. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 276
INSPECTION AND AUDIT ..................................................... 277
Procurement of Goods and Services . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 277
Resident Inspection Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 277
Guidance for Safeguards Upgrade Rule. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 277
Automatic Data Processing. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 277
Review of Semiscale Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 277
Levels of Review of a Significant Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 278
FUNDING AND BUDGET MATTERS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 278
Contracting and Reimbursable Work ......................................... 278
AUTOMATED CONTROL OF DOCUMENTS ..................................... 278
PHYSICAL FACILITIES ........................................................ 279
NRC LICENSE FEES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .... 279
NATIONAL EMERGENCY PREPAREDNESS ..................................... 280
NRC RESOURCES ............................................................. 280
NRC FINANCIAL STATEMENTS ............................ " ................. 281

Appendices
Appendix I-NRC ORGANIZATION ............................................. 283
Appendix 2-NRC COMMITTEES AND BOARDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 286
Appendix 3-PUBLIC DOCUMENT ROOMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 289
Appendix 4-REGULATIONS AND AMENDMENTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 294
Appendix 5-REGULATORY GUIDES ............................................ 298
Appendix 6-NUCLEAR ELECTRIC GENERATING UNITS IN OPERATION, UNDERCONSTRUCTION OR PLANNED ................................... , 301

Index ................................................................ 312

xi

Statutory Reporting Requirements Addressed
Energy Reorganization Act of 1974, as Amended
Section 307(c) directs the Commission to include in its Annual Report statements anddescriptions concerning:" ... the short-range and long-range goals, priorities, and plans of the Commission as
they relate to the benefits, costs, and risks of nuclear power." (See Chapter 1 for overall
statement. Specific goals conct~rning nuclear power reactors are also discussed in
Chapters 2 and 3; fuel cycle in Chapter 4; safeguards, Chapter 5; wastes, Chapter 6; inspection and enforcement, Chapter 7; emergency response planning, Chapter 8; nuclear
nonproliferation, Chapter 9; standards, Chapter 10; and research and risk assessment,
Chapter 11.)" ... the Commission's activities and findings in the following areas"(1) insuring the safe design of nuclear power plants and other licensed facilities ... "(For reactors, see Chapters 2, 3, 10 and 11; materials facilities, devices andtransportation packages, Chapters 4, 10 and 11; waste facilities, Chapters 6 and
10.)"(2) investigating abnormal occurrences and defects in nuclear power plants andother licensed facilities ... " (See Chapters 2, 3, 7 and 8.)"(3) safeguarding special nuclear materials at all stages of the nuclear fuel cycle ... "(See Chapters 5, 10 and 11.)"(4) investigating suspected, attempted, or actual thefts of special nuclear materials
in the licensed sector and developing contingency plans for dealing with such incidents ... " (Chapters 5, 7 and 10.)"(5) insuring the safe, permanent disposal of high-level radioactive wastes throughthe licensing of nuclear activities and facilities ... " (See Chapter 6.)"(6) protecting the public against the hazards of low-level radioactive emissions fromlicensed nuclear activities and facilities ... " (See Chapters 2, 3, 4 and 10.)
Section 205 requires development of "a long-term plan for projects for the development of new or improved safety systems for nuclear power plants" and an annual updating of the plan. (See Chapter 11.)
Section 209 requires the Commission to include in each Annual Report a chapter
describing the status of NRC's domestic safeguards program. (See Chapter 5.)
Section 210 directs the Commission to submit "a plan prOViding for the specification
and analysis of unresolved safety issues relating to nuclear reactors," and to include progress reports in the Annual Report thereafter concerning corrective actions. (See Chapter
3.)

Nuclear Nonproliferation Act of 1978
Section 602 requires annual reports by the Commission and the Department ofEnergy to "include views and recommendations regarding the policies and actions of theUnited States to prevent proliferation which are the statutory responsibility of those agencies ... " (See Chapter 9.)

Atomic Energy Act of 1954, as Amended
Section 170 i directs the Commission to report annually on indemnity operations implementing the Price-Anderson Act which provides a system to pay public liability claimsin the event of a nuclear incident. (See Chapter 3.)

1
Overview

John F. Aheame, who was named NRC
Chainnan in December 1979, testifies at acongressional hearing.

This is the fifth Annual Report of the U.S. Nuclear
Regulatory Commission. It is submitted to the President for transmittal to the Congress as required by
Section 307 (c) of the Energy Reorganization Act of
1974.
This report highlights major NRC activities in fiscal
year 1979 under headings which correspond with thevarious facets of the agency's statutory responsibility.
This introductory chapter presents a brief overview ofthese activities and provides updating on significant
events and actions extending into early 1980.
The accident at Three Mile Island had a profoundeffect on the public, the utilities, the nuclear industry,
and the Nuclear Regulatory Commission. The signs ofchange in the regulatory area are evident throughout
this report as external and internal examinations of
NRC have resulted in new policy directives. The
chapters dealing with reactor regulation, inspectionand enforcement and safety research discuss thesechanges in some detail. While the primary goals of
nuclear regulation-protecting the public health andsafety, safeguarding nuclear materials and facilities,
and preserving environmental values-remain thesame, the means needed to achieve these goals arechanging to reflect the lessons of Three Mile Island.
If there is a common thread revealed in the ongoingreassessments, it is a complacency that has served toundercut the many conscientious efforts to assure
nuclear safety. Current reappraisals must be seen as
opportunities to orient nuclear regulation away fromthat complacency and toward an outlook befitting atechnology that combines remote possibilities of fatal
accidents or catastrophes with substantial day-to-day
benefits. It is the Commission's philosophy that
nuclear regulation must reflect a continuing commitment to come to grips with the realities of nuclear
technology and its relationship to those who control it,to those who work with it, to those who live near it,
and to the public at large.

As part of this commitment, the Commission is in
the process of reappraising its priorities. It has decided
to give explicit guidance to the staff for use in preparing plans, budgets, and programs over the next few
years. As this Annual Report was being prepared., theCommission-for the first time-was developing a
Policy, Planning, and Program Guidance document.
The document includes the Commission's direction as
to which regulatory areas need greater emphasis inplanning for future agency activities such as:• To define more clearly the level of protection of
the public health and safety that the Commission
believes is adequate based on statutes, public input, and NRC's subjective and quantitativeevaluations.
• To increase efforts to describe to the public therisks of nuclear activities and the uncertainties in
the judgments of risk.
• To regulate nuclear activities in a manner toachieve and maintain adequate protection of
public health and safety. Licensees who cannot
do this will not be permitted to 9perate.
• To give priority in reactor regulation, in terms ofresources and schedules, to those activities thatare expected to have the greatest effect on reduction of risks to the public health and safety. Firstpriority will be assigned to operating facilities.
Priorities of NRC activities involving thoseresources not engaged in assuring adequate levelsof protection for operating facilities will beassigned according to risk reduction potentials.
• To organize and plan a waste management program to achieve in a timely fashion the ultimate
objective set forth in the President's Policy Statement of February 12, 1980 on waste management. The NRC waste management program is
critical to the success of this urgent national task.

2

NRC Commissioners testify at one of numerous Congressional
hearings conducted on the TMI accident. From left: Commis·
sioners John F. Aheame (named Chairman in December) and• To emphasize prompt and vigorous enforcement
in dealing with licensees who are unable or unwilling to comply with NRC requirements.
• To require, in any consideration of regulatory
costs to licensees and their customers, that cost
factors be set forth explicitly and applied withpublic health and safety being the paramountconsideration.
• To consider, in determining the adequacy of
public protection, the health and safety implications of not operating a facility as well as thepotential radiological or other hazards of its
operation.
• To increase emphasis on minimizing the consequences of possible accidents, theft or diversion of
nuclear materials, and sabotage or other illegal
acts.
• To license or permit continued operation of a
nuclear facility only when the NRC is confident
that, after termination of the license, there will
be adequate protection of the public health andsafety from potential hazards of the decommissioned facility itself and from wastes associatedwith it.
• To continue a research program whose objectives
are (1) to assist in determining adequate levels ofprotection of the public health and safety and (2)to explore ways to achieve improved levels of protection. The research program should not include
items that should be supported exclusively by theprivate sector. The research program must be
focused on identifiable needs, and its relevance tothe agency's regulatory mission must be the paramount basis for the program.

Richard T. Kennedy; Chairman Joseph M. Hendrie; and Commissioners Victor Gilinsky and Peter A. Bradford.

These policy and planning guidance statements
form the basis for more detailed policy and budget
guidance on each of the important NRC programs.

HIGHLIGHTS AND UPDATES
Accident at Three Mile Island
The accident that occurred on March 28, 1979 atThree Mile Island (TMI) Unit 2 was a traumatic eventfor the American public-especially for the public living near the facility-as well as for the licensee andother utilities; nuclear plant designers, manufacturers
and suppliers; local, State, and Federal authorities
responding to the emergency; and the Nuclear
Regulatory Commission. The extent of the accident's
impact on all of these and on the future of commercial
nuclear activity may not be assessable for a long time,
but it is certain that it is and will be a profound andlasting one. As serious as the event was, major investigations agreed that releases of radioactive
material from the facility were low throughout thecourse of the accident.
Chapter 2 of this annual report, which is devoted
entirely to the TMI accident, includes a narrative of
the events of March 28-April 1; actions taken and investigations made by NRC in the aftermath of the accident up to the end of 1979, with conclusions andrecommendations; and a full account of the findings
and recommendations of the President's Commission
on the accident and the NRC's response to each ofthem.
The President's Commission on the TMI accident
was established on April 11, 1979. President Carter3

The reports of the President's Commission on the
TMI accident, NRC's Special Inquiry Group, and therecent five-year evaluation of the NRC by the General
Accounting Office ("The Nuclear Regulatory Commission: More Aggressive Leadership Needed") stressed
the need to improve the overall managerial functionsof the Commission as a means of improving reactor
safety. The Commission did not agree with the recommendation that the NRC be made an Executive
Branch agency headed by a single administrator.

In addition to the formulation of explicit policy,
planning, and program guidance mentioned earlier,
the Commission has taken or is taking the following
steps to provide more effective agency management:• The Commission continues to pursue the consolidation of all NRC offices at a single location asa means of increasing effectiveness, as recommended by Congressional committees and various
investigatory bodies.
• The Commission is moving to correct the significant organization and management weaknesses
that were revealed by the TMI accident andsubsequent investigations. These actions include
clarifying the role of the Executive Director forOperations, making the Chairman solely responsible for emergency response, giving increased attention to human factors in nuclear regulation,
developing mechanisms to assure more effective
use of advice offered by the Advisory Committee
on Reactor Safeguards, providing a staff
dedicated solely to the evaluation of operating experience, and providing for a more effective roleof the public in reactor licensing.

An audience of NRC staff members and the public listens intently
during a briefing at Commission offices concerning developments

at the Three Mile Island site seven days after the beginning of theaccident.

charged the Commission with investigating the accident and reporting to him within six months withrecommendations based on its findings.
The NRC also chartered its own inquiry into the accident, under independent directorship, the results ofwhich were published at the end of January 1980. In
general, the conclusions and recommendations of this
NRC Special Inquiry Group were consistent with, butmore detailed than, those of the President's Commission. The Special Inquiry Group's report was still
under review by the Commission in early 1980.

NRC Organization and Management

4• The Commission is considering appropriatedelegations of authority to the NRC staff thatwould permit increased concentration by theCommissioners on matters of overriding
significance.
As this report was in final preparation, the President
sent to the Congress an NRC reorganization plan
designed to improve agency management by, among
other actions, strengthening the role of the Chairman.

Reactor Safety
In July 1979, an NRC task force-brought together
to develop a systematic NRC response to the several inquiries and investigations of the TMI accident-recommended a number of short-term actions
to improve power reactor safety (see Chapter 2,"TMI-2 Lessons Learned Task Force
New requirements were issued to all operating reactor
licensees with the objective of completing the changes
by January 1, 1980. Some licensees had difficulty
meeting the deadline because of delays in obtaining
necessary equipment. Thus, while most licensees had
made significant progress by the end of 1979, the NRC
took further action by issuing orders making continued
operation of the reactor(s) in question contingent upon
all changes being implemented by January 31, 1980.
Extensions of the deadline were permitted only when alicensee could show that the needed equipment could
not be obtained within the time frame or that areasonable delay would alleviate severe power supply
problems. In no case, however, will plants be allowed
to operate beyond June·1, 1980 without completion of
the changes.
By the end of 1979, the findings of the various investigations and studies of the TMI accident were used
U).

This mock-up of the TMI-2 control room was constructed ofprecisely measured photographic enlargements of the real control
room by a behavioral science research firm in Arlington, Va. The

by the staff to draft a proposed program of work,"NRC Action Plans Developed as a Result of theTMI-2 Accident" (NUREG-0660). This draft "TMI
Action Plan" contains schedules and resource requirements for more than 100 tasks to provide a higher
level of protection of the public health and safety.
Although many of the proposed tasks are to be completed in 1980, a significant number are multi-year
projects that may extend into the mid-1980's.
No operating license was issued for a nuclear power
plant during fiscal year 1979, and the need to devote
licensing staff resources to TMI-related issues applicable to operating reactors, and subsequently to apply the lessons learned to new plants not yet in operation, resulted in a licensing hiatus that extended into
early 1980. Further, because the TMI accident raised
significant questions concerning the effectiveness of
NRC regulations and. practices in assuring adequate
protection to the public, the Commissioners announced in October that new construction permits,
limited work authorizations, or operating licenses
would be issued only after careful review by the Commission itself. *Certain NRC actions also are intended to improve
the technical and managerial competence of reactor
licensees and the quality of the human element inachieving the safe operation of nuclear reactors.
Essentially, these actions are concerned with the interaction between man and machine-the "human,. On February 28, 1980, the Commissioners approved issuance of alicense to the Tennessee Valley Authority authorizing the loadingof fuel in its Sequoyah Nuclear Power Plant Unit I, and performance of low-power testing under specified conditions. Several
other similar actions were under consideration.

model was used in probing aspects of "human factors engineering"in the investigation by the NRC's independent Special Inquiry
Group.

=======================================================5
factors" whose critical importance was revealed in the
TMI accident. Proposed projects range fromestablishing new requirements for staffing and man~ning control rooms and for training and qualificationof reactor operators and their supervisors to developing and procuring training and engineering
simulators. In cooperation with the nuclear industry,
the staff also proposed to study and identify means for
improving control room design.

Emergency Preparedness
The accident demonstrated that emergency
preparedness both on-site and off-site should be considered comparable in importance to other elements of
the "defense-in-depth" approach to nuclear safety,
and that substantial work must be done in emergency
preparedness, particularly with respect to off-site
preparedness to deal with the aftermath of an
accident.
.
Soon after the accident, Chairman Joseph Hendriewrote to the Governors of applicable States urging thedevelopment and testing of emergency plans aroundthe sites of nuclear power reactors in operation orunder construction. Most of the State and local
authorities have begun to move ahead with development or refinement of their plans.
In December 1979, the Commission proposed newrules that would require upgraded emergency plans inareas near nuclear power plants and concurrence by
the NRC and the Federal Emergency Management
Agency (FEMA) in State and local plans as a conditionof continued operation of existing plants and issuanceof new operating licenses. Both NRC and FEMA are
working with State and local authorities to test andevaluate off-site emergency plans.
From April through July of 1979, a continuous
watch was established in each region and at the NRC
Incident Response Center in Bethesda, Md., to provide for immediate response to any incident or accident. In August, a communications system directlytied into the Operations Center (where 24-hour
coverage is maintained) became operational, thus
relieving the 24-hour duty offices in the Regions.
Other NRC priorities for improving emergency
capabilities include:• Developing NRC, licensee, and State/Local
emergency procedures for all appropriate
f acili ties.
• Instituting for emergency planning purposes azone concept that would establish two concentriczones around each nuclear power plant-the firstfor plume exposure pathway and the second forthe ingestive exposure pathway (milk andagricultural products).
• Requiring additional instrumentation thatwould: (1) provide more precise information on
the status of key equipment during an accident,

Several hundred State and local employees assigned to radiologicalemergency response teams have received training in the Radiological Emergency Response Operations course conducted at
DOE's Nevada Test Site.

and (2) expand the means for measuring radioactive releases around major nuclear facilities.
• Upgrading NRC, licensee, and local communications facilities for prompt NRC response to
emergencies.

Analysis of Operating Data
The TMI accident revealed a need for NRC to put
greater effort into systematically analyzing operational data from nuclear power plants to detect trends
that would better enable all concerned to identify safety problems and take action before they cause
accidents.
In July 1979, the NRC established an Office forAnalysis and Evaluation of Operational Data to conduct systematic and rigorous assessments of licensee
operating experience. The new office will analyze andevaluate operational data associated with all NRC
licensed activities; develop formal NRC guidance on
the collection, evaluation, and feedback of operational
data; and take cognizance of the similar efforts of
NRC program offices, industry organizations, andforeign countries. NRC reactor licensees will be required to conduct analyses of their operating experience and disseminate the results.
Within the industry, two new organizations, theNuclear Safety Analysis Center and the Institute of
Nuclear Power Operations have been created tosystematically review plant operating experience.

6

These organizations will develop and implement programs designed to ensure a high quality of operation in
nuclear power plants. The NRC's exchange arrangements with other countries and international
nuclear organizations provide it with data on
operating experience of overseas reactors, many ofwhich are of U.S. design.

Inspection and. Enforcement
For some time, NRC had realized that greater NRC
presence is needed at major licensed facilities. The
resident inspector program provides this increased
presence at nuclear power plants and other selectedfacilities. The resident inspectors conduct frequent.
direct observation of licensee activities, thereby relying less on the records and reports which were theprincipal sources of information in the past.
The resident program, approved by the Commission
in 1977, has been expanded to entail the placing of atleast two resident inspectors at each operating nuclear
facility site in fiscal year 1980. By December 31, 1979,
a total of 60 inspectors had been deployed at 45
nuclear power stations and three fuel facilities. By
June 1980, each site with a reactor in operation or
about to go into operation (as well as a substantial
number of reactor construction sites) will have at least
one resident inspector.
NRC is currently examining its enforcement policy
and practices. The Commission is awaiting Congressional action on a proposal to increase the civil penalty
authority from a maximum of $5,000 to $100,000 per
violation; the higher authority is more in line with thatavailable to other agencies with public health andsafety missions. The Commission is also preparing a
restatement of enforcement policy that would implement the new authority and provide clear guidance tothe staff.

An NRC reactor inspeCtor (right) at work.

Radiation Protection
Significant steps are being taken in improving ourunderstanding of the potential health effects from exposure to low-level radiation. Holding radiation exposures as low as reasonably achievable under normal
conditions is a fundamental objective of NRC's radiation protection activities. NRC is taking the following
steps to achieve this goal:• Participating in the President's Radiation Policy
Council to improve the coordination of Federal
radiation protection activities.
• Working closely with the Environmental Protection Agency and other Federal agencies todevelop improved standards for controlling occupational exposures. Cooperation with EPA includes joint hearings on occupational exposure
standards, coordination of research programs,
and a study on the health effects of low-level
radiation.
• Improving NRC radiation protection criteria forthe adequacy of licensee health physics programs,
and conducting in-depth radiological safety
evaluations at every operating reactor.
• Working with the National Institute of Occupational Safety and Health to establish a TMI
worker registry that could be used for follow-up
health studies.

Research
As a direct result of the TMI accident, NRC's
research emphasis is being shifted from large-break,loss-of-coolant accidents to small-break LOCA's andrelated transients. Research effort is increasing in theareas of verification of computer codes, fuel behavior
under accident conditions, monitoring of radioactive
releases, emergency response planning, and risk assessment. Research has also been initiated in areas notadequately considered in the past-fuel damage andits effects, core melt, and containment integrity. Also,a n~w study has begun into simulators, control rooms,
andl human factors in nuclear power plant operations.
llhe accident has also underscored the need to apply
the ifault-tree/event-tree techniques used in the Reactor ISafety Study (WASH-1400) to each operating
plant, in an effort to identify the likelihood and consequences of the accident sequences which are thelargest contributors to risk. An "Interim Reliability
Evaluation Program" to review all operating reactors
over the next few years is already under way.
Ip. areas unrelated to TMI, NRC research on seismic
anq. structural engineering and code verification wasincreased during the reporting period following the
shutdown of five reactors in early March 1979 becauseof errors in the seismic analyses of the plants. Research
into the development and application of risk assessment techniques has allowed the Commission to iden-====~~~====~====================================7tify and concentrate on the resolution of those genericissues that involve the highest risk to the public healthand safety. Research has also increased substantially in
waste management to provide technical data andmethods needed to implement regulation being
developed in that area.
In general, future research planning at NRC will bebased on (1) assisting in the determination of adequatelevels of public health and safety protection and (2) exploring ways to achieve improved protection levels.
NRC research must be balanced among confirmationof existing practices, exploration of areas where newconcerns may exist or where existing regulatory approaches may be inadequate, and examination of concepts for improving safety. The program must also becapable of some effort in areas with potential longrange benefits as well as work addressing more immediate goals and needs arising from current NRC activities.

Low-Level Waste. Major emphasis is being placedon developing a regulation for the disposal of low-level
waste in land facilities.
The importance of assuring adequate regulation of
the disposal of low-level radioactive wastes was
dramatized during the past year by the temporary
closure or restriction on operations of each of the country's three existing commercial waste disposal
facilities. The States of Nevada, Washington, andSouth Carolina, which regulate these facilities underagreements with the NRC, closed or restricted operation of the three sites because of deficiencies in the
packaging, transport, or disposal of the wastes being
received. Extended closures could have resulted in curtailment of nuclear medicine services. All of these sites
have subsequently been reopened.

Waste Management
The NRC waste management program is critical to
the successful resolution of the urgent national nuclear
waste problem. NRC is organizing and planning its
program to be consistent with the requirements set
forth in the President's Policy Statement of February
12, 1980, on nuclear waste management.
The NRC has intensified its efforts over the last year
in preparing or modifying regulations covering all
types of nuclear wastes including high- and low-level
wastes and uranium mill tailings.
High.Level Waste. The accumulation of spent fuelcontinues to lead utilities to seek expanded storage
capacity of pools at reactor sites and occasionally toship irradiated fuel from site to site to utilize unused
capacity.
It is estimated that these pools will be filled by 1983.
Although sites away from reactors will be used to store
excess spent fuel, the President, NRC and DOE consider both these sites and the reactor pools only interim
measures before the ultimate solution is provided bypermanent geologic disposal. NRC is developing ageneral regulation (10 CFR Part 60) on the disposal ofhigh-level radioactive wastes in geologic repositories
which is expected to be published in two parts: theprocedural requirements and the technical requirements. The procedural portion was published asa proposed rule for public comment in December
1979, and the technical part will be published in 1980
as an advance notice of proposed rulemaking. Operation of the first repository could begin in the
mid-1990's.

Governors of the three States with low-level radioactive waste
facilities met with the Commission in November to discuss deli.
ciences in the packaging, transport and disposal of the wastes being
received. Shown, from left, are Governors Oixy Lee Ray,
Washington, and Richard W. Riley, South Carolina; NRC Chairman Joseph M. Hendrie (standing) and Governor Robert List,
Nevada.

Uranium Mill Tailings. The NRC is proposing to require the disposal of uranium mill tailings
underground instead of continuing to permit them tobe piled on the surface with virtually no controls. A
draft generic environmental impact statement on
uranium milling, emphasizing tailings management,
was published for public comment in April 1979,
followed by a proposed regulation in August. The
regulation is expected to be in place in late 1980.
Meanwhile, all current licenses under NRC jurisdiction have been upgraded appropriately, and NRC isworking with the Agreement States to upgrade theirprograms accordingly. The Uranium Mill Tailings
Radiation Control Act of 1978 requires that, by late1981, the minimum standards in Agreement States beequivalent to those of NRC regulations. The faih"lre of
the tailings dam at Church Rock, New Mexico, in July
1979 illustrates the need for more stringent regulatory
control over tailings.

8

NRC inspector examines low-level rad.ioacti~~ waste con~iners
before disposal at Beatty, Nevada bunal faclhty. Inspections wereincreased at receiving areas at behest of the Governors of Nevada,South Carolina and Washington.

Transportation
Substantial public concern over the transport ofnuclear products (especially spent fuel) has prompted
some Federal, State and local authorities to enact orconsider restrictions affecting the highway transportation of radioactive materials. In April 1978, DOT
published an opinion in the Federal Register concluding that it had the authority under the Hazardous
Materials Transportation Act (HMTA) to preempt
State and local routing requirements that are inconsistent with DOT regulations.
The NRC has amended its rules to impose DOT
regulations on NRC's licensees. This action is expected
to enhance the NRC's inspection and enforcement effort.
In 1979, the Commission also took two other actions
which contribute to the regulatory framework fortransportation of radioactive waste: it issued guidanceon methods of packaging and it issued an interim rule
requiring safeguards measures for spent fuel shipments
and NRC approval of routes for shipping spent fuel. A
supplemental draft generic environmental impact
statement on transportation should be published in
1980.

Domestic Safeguards
Under the Atomic Energy Act of 1954 and theEnergy Reorganization Act of 1974, the NRC is
responsible for the regulation of safeguards provided
by certain of its licensees. NRC safeguards regulatory
programs share the common goal of assuring thatlicensed activities do not pose undue risk to the publichealth and safety and are not inimical to the common
defense and security. The NRC safeguards objective isto require the implementation of measures designed toprevent, deter, detect and respond to (1) the
unauthorized possession, theft, diversion or use of
special nuclear material (SNM) a~d (2) t?~ s.abotage of
nuclear facilities and transportatIon actiVItIes.
NRC currently exercises safeguards regulatory control over 19 fuel cycle facilities that are authorized topossess formula quantities of highly enriched uranium
or plutonium, transportation activities involving spent
fuel or formula quantities of highly enriched uranium
or plutonium (about 20 shipments per month), 70
power reactors and 71 non-power reactors. NRC also
has safeguards responsibilities for other facilities
which possess significant quantities of low enriched'uranium as well as numerous small facilities thatpossess and ship SNM.
In November 1979, a physical security upgrade rule
was published in final form. It became effective inMarch 1980. The rule indicates performance standards
to be met by licensees in protecting formula quantities
of SNM and presents specific statements about thekinds of threats, from insiders and outsiders, that theirsafeguards should be able to withstand. Certain nonpower reactors are temporarily excepted from the new
upgrade rule, but will be subject to interim requirements. All non-power reactors are now subject tospecial protection requirements.
As a result of an excessive inventory difference inAugust 1979, a fuel cycle facility operated by Nuclear
Fuel Services at Erwin, Tenn., was ordered to be shut
down for investigation and reinventory. While the inventory difference was not fully reconciled, the facility was allowed to resume operations in January 1980
subject to the implementation of additional physicalprotection and material control and accounting
measures, with a study of potential alternative
measures for improving accountability to be completed by DOE within one year. In addition, thelicense will be required to conduct a special reinventory in the event that an inventory difference is found
to exceed 1.0 percent of throughput. Further, a plant
shutdown would be required if the special reinventory
does not reduce the inventory difference to below 1.5
percent of throughput. These new limits, which areless restrictive than the former limits, are considered tobe representative of the level reasonably achievable forthe process.
In 1979, NRC consolidated responsibility for integrating and coordinating the overall NRC
safeguards program in the Office of Nuclear Material
Safety and Safeguards (NMSS). NMSS coordinates
those safeguards activities pertinent to reactors with
the Office of Nuclear Reactor Regulation.
A detailed report on the status of domestic
safeguards during fiscal year 1978 was sent to Congress on February 1, 1979, as required by Public Law
95-601, amending Section 209 of the Energy

~~~~~==~==~============~================9Reorganization Act. The follow-up report for fiscal
year 1979 is presented in Chapter 5 of this Annual
Report.

Exports and International Safeguards
Under provisions of the Atomic Energy Act of 1954
and the Nuclear Non-Proliferation Act of 1978, the
NRC ensures, through licensing, that effective U.S.
controls are applied to the export and import ofnuclear materials, equipment and facilities. It is also
NRC policy to support the reliability of the U.S. inmeeting its supply commitments to nations which
adhere to effective non-proliferation policies by implementing procedures that facilitate the timely processing of export licenses. In addition to exercising its
direct licensing authority, the NRC consults with the
Departments of Commerce and Energy on nuclear
export-related functions under their authority.
During fiscal year 1979, the NRC issued 678 nuclear
export licenses and amendments, of which 154 weremajor licenses; provided views to DOE on 13 requests for approval of retransfers of U.S.-origin spent
fuel to other countries for reprocessing; and consulted
with DOE on several cases involving the export oftechnology associated with the production of special
nuclear material outside the United States.

Spent fuel cask designed for highway transport is shown being
placed on a protected truck flatbed. The cask, 18 feet lon~ andThe Commission recently confronted the question of
its authority and responsibility to consider the effects
of reactor exports on the health, safety and environment in recipient countries. The question arose in thecontext of the Commission's consideration of a controversiallicense application to export a reactor to thePhilippines. On January 29, 1980, the Commission
decided to limit its review in the Philippines and otherreactor export cases to health, safety and environmental factors affecting the global commons or the territory of the United States, and the relationship ofthese effects to the common defense and security of the
U.S. Consideration of local impacts, including effects
on any U.S. citizens located there, would continue tobe the sole responsibility of foreign recipient governments.

THE NRC COMMITMENT
Despite the problems and issues to be resolved,
nuclear electric power is still acknowledged to be an
important element in the nation's energy strategy forsome time to come. The President has stated that
"nuclear power has a future in the United States-it isan option that we must keep open. I call on the utilities
and their suppliers, the NRC, the Executive Departments and agencies, and the State and local governments to assure that the future is a safe one."fo~r feet in diameter, weighs 50,000 pounds when fully loaded
With spent fuel assemblies.

10

It is clear that the next few years will see many
significant changes in the NRC's efforts to assure thatthe public health and safety is adequately protectedfrom the potential risks of nuclear power plants. Thechanges will be widespread-in the NRC organizationand its relations with Federal, State, and local
authorities; the regulations; the design of nuclear
plants; the utilities' mode of operation; and theregulatory process itself.
Neither the President's Commission nor the NRC's
own Special Inquiry Group was able to define an explicit standard of nuclear safety-or, more simply, toanswer the question, "How safe is safe enough?" It isquite clear that society itself must ultimately provide

the answer as to what is acceptable. All that has happened during the past year has served to confirm the
proposition that the part nuclear energy will play in
the U.S. energy strategy is directly and inevitably
linked to the public's perception of nuclear safety.
As the agency responsible for nuclear regulation, the
NRC must play the fundamental role leading to theproper determination of what is an adequate level ofprotection. The NRC must bring its management andtechnical expertise to bear in assuring that theregulated industry achieves and maintains that protection. The NRC is fully committed to meeting thischallenge.

2
Accident atThree Mile Island
The accident at Three Mile Island in March
was the focus of nearly all NRC activity
in 1979.

Despite the fact that no one was killed and nophysical injuries were sustained among the generalpublic because of it, the accident at the Three Mile
Island Nuclear (TMI) Station Unit 2 is unquestionably
the most serious in the history of commercial nuclear
power. It is also the most intensively studied and extensively reported incident in that history. This
chapter can only attempt to cover the major investigative efforts devoted to the accident, only those
whose results were available before the end of 1979
(the NRC's own Special Inquiry Group report was
pending, as were the results of several Congressional
studies), and only the most salient findings and recommendations or actions issuing from them. Other
chapters of this report cover many aspects and effectsof the TMI accident in connection with the particular
NRC activities under discussion. These references arecited in the Index under "Three Mile Island accident."
The full reports of the various NRC investigations
and other documents cited in this chapter are available
from the GPO Sales Program, Division of Technical
Information and Document Control, U.S. Nuclear
Regulatory Commission, Washington, D.C. 20555,
and from the National Technical Information Service,
Springfield, Va., 22161. The titles and catalogue
numbers are listed in the box below. The report of the
President's Commission on the Accident at Three Mile
Island, which is discussed at length, is available fromthe U.S. Government Printing Office.

WHAT HAPPENED
Located in Dauphin County, Pa., about 10 miles
southeast of Harrisburg on an island in the Susquehanna River, the Three Mile Island Nuclear Station (TMI)
consists of two pressurized water reactors andassociated equipment, each one with two large steam
generators and each employing two 370-ft. cooling
towers-part of the system which condenses the steam
after it has passed through the turbines to generate

electricity. The utility licensed to operate the facility is
the Metropolitan Edison Company, a subsidiary ofGeneral Public Utilities, Inc., of New Jersey. Unit 1 at
TMI was licensed for operation in 1974, at a net
capacity of 819 MWe; Unit 2 was licensed in
February 1978 and went into commercial operation in
December 1978. Each unit has its own reactor containment building, control room and auxiliary building. Each containment building houses a reactor, apressurizer, and two steam generators; the turbine
and electric generator are outside the containment.

NRC REPORTS ON TMICITED IN THIS CHAPTER
NUREG-0558: "Population Dose and Health Impact of
the Accident at the Three Mile Island
Nuclear Station"
NUREG-0560: "Staff Report on the Generic Assessmentof Feedwater Transients in Pressurized
Water Reactors Designed by the Babcock& Wilcox Company"
NUREG-0578: "TMI-2 Lessons Learned Task Force:
Status Report and Short-Term Recommendations"
NUREG-0585: "TMI-2 Lessons Learned Task Force:
Final Recommendations"
NUREG-0596: "The Non-Radiological Consequences to
the Aquatic Biota and Fisheries of theSusquehanna River from the 1979 Accident at Three Mile Island Nuclear Station"
NUREG-0600: "Investigation into the March 28, 1979
Three Mile Island Accident by [NRC] Office of Inspection and Enforcement"

12

The Three Mile Island Nuclear Station. The four large towers cool
the steam generator water used in the production of electricity by
the two reactor units at the station, located in the cylindrical

domed structure shown at right center in the photo. The Unit 2
reactor, scene of the accident, is housed in the cylindrical containment building farthest to the right.

Wednesday-March 28

pressure of the water to rise and a "pressurizer relief
valve" to open. The reactor automatically shut down
in response to the increase in primary coolant pressure.
This reactor "scram" took place eight seconds after thecondensate pump shut down on the secondary side of
the system. Instantly the output of heat from nuclear
fission in the reactor core was stopped, but a substantial amount of "decay heat" continued. The production of decay heat, like the momentum of a large shipat sea, cannot be ended by turning off the powersource, and it is essential that sufficient primarycoolant and pressure be maintained even after thereactor has shut down.
Through the first seconds of the accident, the performance of the equipment went according to design
and the sequence of responses to the unexpected interruption of heat transfer from primary to secondary
systems was "normal." After the reactor scrammed
and the relief valve lifted, the primary coolant

At about half a minute past 4:00 a.m., on Wednesday, March 28, 1979, a "condensate" pump and themain "feedwater" pumps connected with one of theUnit 2 steam generators shut down, causing an almost
simultaneous and automatic shutdown of the Unit 2
turbine (Unit 1 was shut down at the time for refueling.) The initiating cause of the shutdown is notdefinitely known but may have been an alteration in
the pressure in the feedwater system brought about by
a maintenance procedure taking place at the time. An
unexpected pump shutdown is not unusual or, in itself,
serious. With the feedwater flow stopped, the steam
generators stopped removing heat from the primarysystem, i.e., from the closed system of pressurizedwater which passes through the reactor, carries heat tothe secondary system, and returns to the reactor. The
buildup of heat in the primary system caused the~~=================================================13

pressure dropped back to the point where thepressurizer relief valve was supposed to close, restoringa closed, fully pressurized primary system with coolant
flowing through the reactor core and removing its
decay heat (about 7 percent of its normal operatingheat production). The relief valve did not close. At thissame time, several pumps came on automatically on
the secondary side to restore feedwater flow andremove heat through the steam generators. This action
was thwarted by closed valves, a condition which was
not corrected until eight minutes into the accident.
Because the pressurizer relief valve was stuck open,
the pressure in the primary system did not level off at
the proper point but continued to decrease. As thepressure of the coolant goes down so does its boilingpoint, and the danger arises that it may begin to turninto steam. Since steam cannot carry off decay heat effectively, the primary system could heat up todangerous levels. When the pressure had decreased toabout 75 percent of normal, an emergency core cooling system (ECCS) automatically came on, injecting
cold water under high pressure into the reactor.

Believing that the pressurizer relief 'valve was closed
and seeing the level of coolant in the pressurizer risewith the injection of ECCS water into the reactor, theoperators in the control room feared that thepressurizer would fill up with coolant and the system
would lose the pressurizing bubble of steam that isnormally maintained at the top of the pressurizer.
Consequently, they shut off one ECCS pump andthrottled back the ECCS flow from the other pump into the reactor. Ordinarily the level of coolant in thepressurizer is an accurate indicator of the volume ofcoolant in the entire primary system, so the operators
were confident that the system was full, the reactor
core was covered, and the heavy injection of ECCS
coolant was unnecessary and was, in fact, making thesystem too full. As the four licensee personnel then present in the control room later testified, they were notaware that the level of coolant in the pressurizer is notnecessarily an index to the amount and level of coolant
throughout the system. As it happened, the drop inpressure following the failure of the relief valve toclose and the failure of the auxiliary feed water

The major buildings making up the Three Mile Island Nuclear Station Unit 2 are shown, including the Epicor-II building at rightwhich houses the system used to decontaminate the radioactive

water held in the auxiliary building tanks. The containmentbuilding at the left houses the Unit 2 reactor, pressurizer andsteam generators. The Unit 1 building is at the far right.

14

allowed the coolant going out of the core to boil, andsteam voids or bubbles had formed within the primarysystem between the reactor core and the coolant in thepressurizer. Under such conditions, the level of coolant
in the pressurizer would not disclose the amount of
coolant in the primary system as a whole.
The pressurizer relief valve remained open for about
two hours and 20 minutes, permitting the escape fromthe primary coolant system of more than 30,000
gallons of slightly radioactive water. Early in the accident, the operators were also letting coolant out
through a "letdown" system, in the belief that thesystem was close to filling up. In fact, more coolant
was leaving the primary system than coming into it,
and this led eventually to "uncovering" of the upper
portion of the reactor core resulting in sharp increases
in temperature, damage to the fuel rods and releases ofradioactive fission products. Just how extensive thedamage was to the core and fuel cannot be determineduntil technicians are able to open the reactor vessel.
Estimates of the damage are based on analyses ofsamplings taken from the atmosphere inside and from
coolant standing on the floor of the containment, andthey tend to indicate extensive damage to the fuel.
At 4:08 a.m., a sump pump came on automaticallyand began moving the slightly radioactive
coolant-which had come down from the drain pipe
for the relief valve and from the letdown system-into

sump tanks located in the Unit 2 auxiliary building. It
was at this point that radioactive material first left thecontainment; some of it was eventually vented to theoutside air (though the more serious releases camelater). At 4: 11 a.m., the reactor building sump
overflowed. Some minutes later the control room crew
was apprised of this and, at 4:39, turned off the sump
pumps in the containment. By that time, somethingover 8,000 gallons of water had been pumped to tanks
in the auxiliary building, which was not sealed offfrom the outside air as the containment building was.
At 4:50 a.m., the superintendent of technical support
for Unit 2 arrived, but he too found a situation he had
never experienced: a high level of primary coolant in
the pressurizer but low pressure in the coolant system.
At 5:14 a.m., reacting to vibrations in the fourpumps circulating coolant through the reactor (causedby steam in the coolant), the operators shut down twoof them. Twenty-seven minutes later, for the same
reason, they shut down the other two, cutting off all
flow of coolant to the reactor core. The expectation at
this stage was that the primary system could now workby "natural circulation" with the coolant heated bydecay heat expanding and moving upward to thesteam generators (whose feedwater was now restored
and would carry off heat from the primary system)
and with the cooler water flowing down and back to
the reactor. The operators did not succeed, however,in establishing natural circulation.

Schematic of the TMI-2 facility.

CONTAINMENT BUILDINGAUXILIARY
BUILDING

TURBINE
BUILDINGGE

RADWASTE
STORAGE TANK

====================================================15
By 6:00 a.m., there was evidence, from radiation
alarms, of radioactive gas in the containment. Primarycoolant continued to escape through the relief valve,
now containing non-condensible radioactive gas andhydrogen generated by a reaction between the zirconium cladding on the overheated sections of the fuelrods and the steam in the system. Finally the relief
valve was sealed when a block valve on the pressurizer
was closed at 6:20 a.m. That action ended the loss of
coolant from the primary system, but the flow of
coolant was not resumed until 6:45, when a reactor
coolant pump was reactivated; vibrations again caused the operators to turn off the pump.
A conference telephone call took place beginningabout 6:00 a.m., involving officials of the licensee
company and a representative of the reactor manufacturer. About the same time, radiation readings at
various points on the island began to show abnormal
increases and instruments in the reactor core registeredabnormally high temperatures. At 6:50 utility officials
publicly declared a "site emergency," a procedure
prescribed in the facility's emergency plans wheneveran event posed the possibility of an "uncontrolled
release" of radiation to the immediate environment.
Local and State authorities were notified of the potential impact on public safety, beginning with the 7:02
a.m. notification of the Pennsylvania Emergency
Management Administration (PEMA). The licensee
tried to contact the NRC Region I office near
Philadelphia starting at 7: 10, but the switchboard

CONDENSATESTORAGE
TANK

there did not open until 7:45. The TMI station
manager arrived on the scene shortly after 7:00 and at
7:24, he declared a "general emergency," signifying asituation with the potential for "serious radiologicalconsequences" for public health and safety.
At 7:45 a.m., the NRC regional office was madeaware of the situation at TMI and established an open
line with the Unit 2 control room within a few
minutes. By 8:00, the NRC headquarters was alerted
and the Operations Center in Bethesda, Md., was activated. The regional office dispatched a first team ofinspectors to the site about this time, and other agen.
cies mobilized in response to communiques from NRC
and State authorities.
Radiation monitoring on and near the island had
begun before 8:00 a.m. and was to broaden and intensify throughout this and subsequent days of the accident. A helicopter engaged by the utility was taking
samples above the plant by midday and another aircraft detailed from the Department of Energy (DOE)
was in action by mid-afternoon of the first day. Fromthe beginning, the level of radioactivity around the
TMI site was in the range of one or two milliremper-hour (thousandths of a rem) on the ground,though readings above the island and at some points
on the plant grounds or just across the river were muchhigher and inside the containment ran up to thousandsof rem-per-hour. The radioactive coolant which had
overflowed the sump tanks in the containmentbuilding was automatically pumped over to the aux-

At left is a schematic drawing of the Three Mile
Island Unit 2 facility. Some of the major components are labeled as follows:(A) REACTOR VESSEL: A cylindrical vessel madeof steel-40 feet high and 8'/2 inches thick-whichcontains the reactor (core and control rods) andthrough which the reactor coolant flows, carrying
heat away from the core to the steam generators.
The TMI-2 reactor contains 177 fuel assemblies
with 208 fuel rods in each assembly.
(B) REACTOR COOLANT PUMP: One of fourpumps which move the reactor coolant through thecore to the steam generators and back to the core ina closed system (the primary system) of what is nor·mally only slightly radioactive water. About one
hour into the TMI accident, the operators shut
down two of these pumps because they werevibrating severely, the result of the steam in the
primary system. Half an hour later they shut down
the other two pumps for the same reason. At thatpoint, damage to the fuel in the core began, caus·
ing releases of radioactive material into the coolant.
(C) PRESSURIZER: A large vessel connected to theprimary system between the reactor and the steam
generators which is normally a little more than half
full of water, with a steam bubble in the upper
portion of the vessel. It is designed to keep thepressure in the reactor coolant relatively constant.
(D) PILOT OPERATED RELIEF VALVE: The
pressurizer relief valve located at the top of thepressurizer and designed to open automaticallywhen primary system pressure rises to a preset level

and it becomes desirable to let off steam. Whenpressure is back to normal, the relief valve is sup·posed to close by itself. At TMI-2 it failed to do so,and reactor coolant flowed through the relief valve
and down to a drain tank on the floor of the con.
tainment building. This valve remained open for
more than two hours.
(E) STEAM GENERATOR: The large vessel inwhich the transfer of heat from the reactor coolant
to the feedwater takes place. The transfer results inthe conversion of the feedwater into steam, as it
flows around tubes carrying the pressurized, core·heated coolant from the reactor. This steam is conveyed to the turbine which powers the electricalgenerator.
(F) CONDENSER: The vessel in which the steamwhich has passed through the turbine is condensed
to a liquid state again. The heat is removedby pipes carrying condenser water whichflows to the cooling towers and back to the condenser.
(G) CONDENSATE PUMP: The pump whichmoves the feedwater (the condensate) from the con·denser to the polisher or demineralizer whichcleanses the water before it flows back to the steam
generator. The TMI accident began at this point in
the feedwater system when plant personnel were
trying to clear a line associated with the polisher
and the condensate pump automatically shut down,
followed by a similar "tripping" of the feedwater
pump and subsequently of the turbine and thereactor.

16

Above is a photo of the conference room at the NRC Operations
Center in Bethesda, Md., taken during the course of the TMI accident. Other rooms of the center are equipped and staffed to gather
and analyze data and maintain seCure communications with NRC
regional offices and the accident site. Numerous technical experts
from the NRC were at the center to inform and advise senior NRC
officials on the Executive Management Team. Personnel fromother l'ederal agencies involved in or assisting with management of
the accident were officed in areas adjoining the Operations Center.
In the foreground, at left, is Lee V. Gossick, NRC Executive
Director for Operations.

iliary building tanks where it again overflowed. Since
the auxiliary building is not isolated from the outsideenvironment, some radioactive gases carried over inthe coolant were vented to the outside. The reactor
containment building was not sealed off from the auxiliary building until about 9:00 a.m., after more thaneight thousand gallons of coolant had been transferred.
This transfer of coolant was not, however, the maincause or source of the release of radioactivity to the environment during the TMI accident. The transfer actually took place prior to any major fuel damage in thereactor. It was between one and two hours following
the turbine trip, when the operators turned off thereactor coolant pumps to save them from vibration
damage, that damage to the nuclear fuel began. For
the next several hours, there was a large temperature
difference between the coolant entering and exiting
the nuclear core, indicating inadequate flow of coolant
through the core. As a result of fuel damage, the concentration of radioactivity in the reactor coolant increased by several orders of magnitude. A flow of thishighly contaminated reactor coolant was maintained
from the primary coolant system through the letdown
system and returned to the primary system via themakeup system. This flow, maintained for several
days following the accident, was necessary to ensure

adequate cooling of the reactor coolant pump bearings. Normally the gases evolving from the reactor
coolant in the letdown and makeup systems are of little
radiological significance. During this period,
however, these gases caused very high radiation levels
inside the auxiliary and fuel-handling buildings andresulted in much higher than normal environmental
releases via the ventilation exhausts from thesebuildings. This flow was the principal pathway by
which radioactivity passed from the damaged reactor
core to the auxiliary building, fuel-handling building,and to the environment.
At about 8:00 a.m., the station superintendent andother officials on the site decided to try again to activate the reactor coolant pumps. After some difficulty, two of the four pumps (one in each loop) wererestarted. By 8:30, there was new coolant entering the
primary system from the ECCS.
At 9: 15, the White House was notified of the accident by the NRC. The team dispatched by NRC
Region I arrived at the site by 10: 15. It was shortly
afterwards that the radiation level in the Unit 2 control room required that personnel there don respiratory
masks. These proved to be a hindrance to clear communication. At 11:00 a.m., all non-essential personnel
were ordered off the island. It was about this time that
the NRC and State radiation protection officials asked
the Department of Energy (DOE) to send a team fromthe Brookhaven National Laboratory to help withradiation monitoring.
About 11:30 there began an attempt to depressurize
the reactor coolant system so as to be able to activate
the low-pressure decay heat removal system. The
pressure, however, remained too high for this purpose
because of the volume of hydrogen gas and steam in
the primary coolant system. Hence, the decay heatremoval sysyem could not be initiated, and the attempt at repressurization was terminated about 3 p.m.
Repressurization began at about 5:30 and was completed at about 6:45.
Sometime around noon, three licensee employees
entered the Unit 2 auxiliary building and found radiation levels of from 50 to 1,000 rem-per-hour; each of
the three incurred radiation doses of 800 millirem. At
1:50 p.m., a hydrogen explosion or "burn" took placein the Unit 2 containment building. Personnel on handlater remembered hearing a thud about this time andthe computer chart showed a sudden pressure surge inthe containment up to 28 pounds-per-square-inch, but
the meaning of the spike on the chart was not immediately recognized.
By evening of the 28th, NRC had 11 people on the
TMI site and a mobile laboratory van for analysis of
the radiation content of environmental samples. A
team from the Brookhaven National Laboratory had
been assisting with the radiation monitoring since
mid-afternoon, as had the aerial survey aircraft from
DOE. About 8:00 p.m., a reactor coolant pump was

==~============~=============================17

activated and coolant flow was established, carrying
heat out of the reactor through one of the steam
generators to the condenser, bypassing the turbine.
The primary system remained essentially in this mode
for a month, until natural circulation was finallyachieved on April 27.

Thursday-March 29
On Thursday morning, a team of seven specialists
from NRC headquarters arrived at the site. At thattime the radiation readings at and near the plant were
not negligible but also were not alarming. No significant iodine releases were detected. These would beconsidered especially hazardous because radioactive
iodine should it enter the human food chain, tends toaccum'ulate in the thyroid and can cause cancer of thatgland. The Congress evinced immediate and urgent
interest in events at the plant: Chairman Hendrie wascalled to explain the situation before the House Subcommittee on Energy and the Environment, andSenators Heinz and Schweiker and Congressmen Ertel
and Goodling-all of Pennsylvania-were briefed by
the utility and the NRC. During the afternoon, somewaste water from the plant was discharged by thelicensee into the Susquehanna River. Because it contained only slightly radioactive material, the release
did not constitute a violation of NRC regulations, but,with all the uncertainties still surrounding the scene at
TMI, the NRC Chairman ordered the discharges stopped. Late in the day, analyses of coolant samples confirmed the presence and showed something of the extent of the core damage that took place during theperiods that the core was uncovered on Wednesday.
(I t was later determined that there had been three
periods when a significant portion of the core was be~
ing cooled by steam rather than fluid coolant.) First
concerns about the presence of a hydrogen bubble in
the reactor vessel arose on Thursday, and the fact that
there had been a hydrogen explosion outside the vesselin the containment building early Wednesday afternoon was brought to light.

Friday-March 30
Friday was the day when it became clear to all concerned that the event was far from over; that radiation
releases from the auxiliary building were not undercontrol and were increasing; that there was a large
gaseous bubble in the reactor vessel which could conceivably expand, forcing the level of coolant below the
top of the core, uncovering it again; that, according tosome analyses and expert judgments, the bubble
might become flammable as oxygen evolving from thedecomposition of water by radiation made its way into
the upper part of the vessel; that radiation was
emanating from the facility in a manner neither planned nor controlled.

Early in the day, reports of a 1200 mr/hr reading
above TMI-2 precipitated serious discussion at the
NRC Operations Center in Bethesda of the possibly
urgent need to evacuate the residents of Goldsboro,
Middletown and other communities and areas aroundthe plant, even out as far as Harrisburg. The fact that
there was a consensus favoring such a recommendation at the Operations Center was relayed to State officials in Pennsylvania, occasioning considerable anxiety and confusion, since the judgment was not sharedby people at the plant site. The NRC position wasclarified when Chairman Hendrie spoke with Governor Thornburgh about 10:00 a.m., and counseledagainst full-scale evacuation of the population, suggesting instead that the G?verno~ recomme~d t?atpeople stay indoors for awhIle untIl the true situatIOn
could be better defined. The Governor did so. About
40 minutes later, President Carter contacted Chairman Hendrie and directed that a senior NRC official
be dispatched to the TMI site as his personal representative; the President also assured that the White House
staff would see to it that an adequate and dependable
communications system would be set up as soon as
possible between the site, the White House and the
NRC. Prior to this, communications between the plant
and the NRC had been unreliable and had even been
lost for a time. The Director of NRC's Office of
Nuclear Reactor Regulation, Harold Denton, left NRC
headquarters for the TMI site with a support staff of 12
to serve as the President's representative and as theprimary NRC official on the scene. Shortly after noon,
Chairman Hendrie indicated to Governor Thornburgh
by telephone that a recommendation by the Governor
that pregnant women and pre-school aged childre~
within five miles of the plant leave the area temporanly was advisable. The Governor made this recommendation soon afterwards.
Discussions and assessments of the possible need for
total evacuation of the population near TMI continued
throughout the day among NRC, other Federal andState officials. About an hour after the former's arrivalat TMI and a first assessment of conditions in andaround the plant, NRR Director Denton and Chairman Hendrie reviewed various possible courses the accident might take-or that licensee personnel mighttake in their effort to gain control of events-and theimplications of each for a judgement on whether andwhen to move people out of the area. Within an hourof their conversation, Chairman Hendrie was in contact with Governor Thornburgh, at which time he advised the Governor that, though the bubble in thereactor vessel could cause trouble later in keeping thecore cooled, there was no appreciable amount of oxygen in it and the chance of a hydrogen explosion suchas took place in the containment on Wednesday was"close to zero." The Chairman also appraised thechance of a core meltdown as being extremely low, but
the possibility of a significant radiation release as heing somewhat higher.

18

Additional contingents of NRC personnel were sent
to TMI during the day and by 4:00 p.m. there were 83
NRC staff people at the site. Other Federal agencies-DOE, EPA, FDAA and others-and State officials responsible for emergency management andradiation protection were also present in force. In apress release issued around 6: 00 p. m., the NRC Chairman declared that there was "no imminent danger of ameltdown" of the reactor core. By 8:30 on Friday
evening, Governor Thornburgh decided, having consulted with NRC officials on the site, to lift the advisory that people within five miles of the plant should
stay indoors but, with NRC concurrence, continued torecommend that pregnant women and young children
leave and/or stay out of the area.
By day's end, there was deep uncertainty among allconcerned as to the potential hazard represented by
the hydrogen bubble in the reactor vessel. National
laboratories and industrial experts, as well as NRC
research personnel, were at work calculating how longit might be until the amount of oxygen finding its way
into the hydrogen bubble would produce a flammablemixture in the upper portion of the vessel. Preliminary
estimates of that time-frame varied. Later on it was
realized that there was no appreciable build-up taking
place because the oxygen resulting from the radiolytic
decomposition of water was combining with freehydrogen in the reactor coolant.

The control room for TMI Unit 2 is shown above. The instruments
and controls are deployed in a U-shaped pattern in a design intended to permit one operator to supervise operations under nor·mal, stable conditions. During abnormal situations. it is expectedthat additional operators would be available to give any needed
assistance.

Saturday-March 31
On Saturday the focus of concern had shifted fromperiodic uncontrolled radiation releases to potential
explosion of the hydrogen in the reactor vessel. Radiation readings were very low everywhere but inside thecontainment. The NRC and other Federal presence atthe site was expanding. The NRC Commissioners
meeting in Washington, D.C., continued discussionsof what changes in the situation might warrant a
recommendation that people leave the TMI area, orwhether such a recommendation should be made immediately, as a precaution. The conditions at TMI-2
were improved in virtually every respect, except for
the hydrogen problem, and the Commissioners were==~~~~~~========~======================~~=19

conscious of the hardships an evacuation would impose upon the population. There was also the matterof range to consider, whether to evacuate out to fivemiles or 10 miles or more, and of how much time
would be available, if core conditions began todeteriorate, before the defensive barriers of the plant
would be breached.
Around noon the NRC Chairman and NRR Director
at the site discussed the situation at length, consideringboth the kinds of events that would signal a need to
begin moving people out and also various means bywhich to reduce the hydrogen hazard. Soon afterwards, estimates were received from various researchteams that the conditions necessary for hydrogen combustion or exploSion in the reactor v~ssel were perhaps
days away, and it appeared that there would be amply
sufficient time to vent the vessel into the containment
or otherwise defuse the danger. In mid-afternoon,
Chairman Hendrie held a press conference at the NRC
Operations Center in Bethesda, Md., at which he affirmed that a precautionary evacuation of the TMI
area was still a possibility, especially if it were decided
to try to force the hydrogen bubble out of the reactor
vessel. Soon after, the Chairman and the Governor
conferred by phone. Responding to the latter's query,the Chairman advised that, since some low-level
releases of radiation were still coming from the auxiliary building, it would be prudent to continue therecommendation on pregnant women and pre-school
aged children and to keep emergency planning personnel and resources in readiness.

Sunday-Aprill
Following a brief meeting with the staff at
Bethesda, Chairman Hendrie left Washington to go to
the TMI site. President Carter was to arrive there in
the early afternoon for a tour of the scene and briefings
on the status of the reactor. During the morning, the
NRC personnel at the site had augmented the radiation monitoring equipment by placing 37 thermoluminescent dosimeters within a 12-mile radius of
the plant. By mid-afternoon the bubble in the reactor
vessel seemed to be dissipating and the system stabilizing, though intense discussion of the evacuation question continued among Commissioners and staff inWashington. Chairman Hendrie communicated thefavorable change in the situation to the group in
Washington and characterized the next phase inmanagement of the accident as a choice between moving at once to activate decay heat removal from thereactor or moving slowly and letting the reactor coolat its own rate.
Reactor cooling was maintained by the action of oneof the main coolant pumps providing the flow throughthe reactor core, and heat removal through one of thesteam generators to the condenser, until about 2:00

Governor Thornburgh and President Carter are escorted into the
TMI UOlt 2 control room by a Metropolitan Edison employee on
Sunday, April 1, 1979.

p.m. on April 27 when the reactor coolant pump was
intentionally shut down and core cooling by natural
circulation was achieved.
A bulletin was transmitted Sunday. afternoon to all
NRC licensees operating reactors of the B&W design tomake an immediate review of plant conditions and to
implement a number of precautionary measures derived from the TMI experience. NRC inspectors were also
sent out to confirm that the prescribed actions weretaken. The bulletin was the first in a series issued by
NRC to licensees as analyses of the TMI accident
revealed both necessary and prudential actions to betaken to prevent recurrence of the event (see "Bulletinsand Orders Task Force," below).
Later in April, licensees for the other nuclear power
plants employing B&W nuclear steam supply systems
indicated that they would voluntarily shut down until
prescribed alterations in design and procedures werecompleted. Confirmatory orders to that effect were
subsequently issued by NRC for several of these units.
By the end of May, "dedicated" telephone lines had
been established between the NRC Operations Centerin Bethesda and 68 of the 70 licensed nuclear power
plants and 14 licensed fuel cycle facilities. The lines
make it possible for operations personnel in thesefacilities to communicate immediately and directlywith members of the NRC's technical staff any time ofthe day or night on any day of the year. The system
also provides for instant communication with anyoneof the five NRC regional offices.
The accident at TMI-2 generated investigations,
reports, findings and recommendations literally too

20

numerous to mention. The balance of this chapter attempts only to describe the major NRC undertakingsin the matter and to cover the findings and recommendations of the commission appointed by President
Carter to conduct an independent investigation of theaccident and its implications, together with NRC's
responses to those recommendations. At the time thisreport was prepared, the work of the NRC Special Inquiry Group-an investigatory body set up by the
NRC under independent directorship-was not yet
complete, nor had the various Congressional reviewers
reported their results.

RADIOLOGICAL CONSEQUENCES TOPERSONS AND THE ENVIRONMENT

Water from the industrial waste treatment systems of the plant(TMI), designed to be non-radioactive, is checked to insure that ishas not been contaminated.

Plotting wind direction to assist in monitoring operations.

Individual and Collective Doses. NRC staff members
participated in an interagency study to determine theindividual and population doses associated with the
TMI accident. The results of the study are presented
and discussed in the NRC report, "Population Doseand Health Impact of the Accident at the Three Mile
Island Nuclear Station" (NUREG-0558). Based on environmental measurements performed during the accident, it was estimated that the maximum individual
off-site whole body dose was about 83 millirem, which
is approximately one-sixth the NRC's allowable maximum whole body dose of 500 millirem-per-year. The
population within 50 miles of the TMI site received an
estimated integrated dose of 3,300 person-rem. This
population dose is expected to result in less than one
additional fatal cancer among the exposed population,in which 325,000 fatal cancers can be expected to occur as a result of other causes.
Radiation doses to licensee employees have also been
estimated. Occupational whole-body doses accumulated from the date of the accident through May
31, 1979, totaled 225 person-rem. These doses werereceived by employees in performing recovery operations after the accident, such as changing filters in thecleanup systems for air leaving the auxiliary building
and fuel-handling building, sampling of air andprimary coolant, decontamination and radioactive
waste processing operations, and routine inspection
and maintenance activites. In the days immediately
following the accident, four persons received exposures exceeding NRC regulatory standards. Two
persons involved in taking a primary-coolant sample
received doses substantially in excess of the standards.
One person received a total body dose of 4.1 rem (theregulatory limit is 3.0 rem), an extremity (finger) doseof 147 rem (the limit is 18.75 rem) and skin dose to the
top of the head of 13 rem (the limit is 7.5 rem). The second person received extremity doses of 54 rem. Twoother persons received whole body exposures of 3.2
rem and 3.1 rem, which are slightly higher than the
NRC limit of 3.0 rem.

~~~~~~==~~~==========~~====================21of the NWS Harrisburg River Forecast Center and
NWS Philadelphia Area Weather Forecast Center.
The staff's estimates of the transport and diffusion of
TMI releases were used in estimating doses for thelocations of dose-rate instrumentation both on the
TMI site and off. Because the magnitude of the release
was unknown during the early stages of the accident,
data from environmental monitors and meteorologicalestimates were used to calculate releases. Atmospherictransport estimates were used to advise evacuation
planners.
In further protection of the environment, the possible non-radiological consequences to the aquatic biota
and recreational fisheries of the Susquehanna River
from the accident at Three Mile Island Nuclear Stationin late March of 1979 were investigated up through thepost-accident period (through June). Data used in theinvestigation incl uded site-specific biological andwater quality information collected by the licenseunder the Environmental Technical Specifications andNational Pollutant Discharge Elimination Systemmonitoring programs, and also information from Stateand Federal agencies, knowledgeable persons, and
Map of the TMI area showing 5-, 10- and 20-mile evacuation
zones.

Environmental Protection at Three Mile Island.
During the accident at Three Mile Island there wasconcern that a core meltdown might occur. This could
have led to the contamination of the groundwater ofthe island and ultimately of the Susquehanna Riverand beyond. The staff developed contingency plans to
mitigate the effects of groundwater contamination byisolating the immediate plant area from the regional
water supplies. The plans provide for blocking
groundwater movement, for withdrawing the potentially contaminated water, and for monitoring andtemporarily storing the contaminated water. Working
with the U.S. Army Corps of Engineers, the NRC staff
formulated a plan to construct a bentonite,,:cement
cut-off wall, dewatering wells, and a pumping system.
Availability of equipment needed to carry out the plan
was verified. Plans for monitoring and on-site storage
were also completed. It did not prove necessary to implement planned isolation of the area.
Another problem encountered in the accident was
the need for the staff to produce estimates of thetransport or diffusion of gaseous releases, in order toplan for possible evacuation of the population and forassessment of the consequences thereof. Theseestimates were made by staff meteorologists assigned
to the NRC Operations Center. Well into the accident,
the staff ascertained that meteorological data wereavailable from the TMI meteorological tower by
remote access and made use of this information. In addition, the staff arranged for National Weather Service (NWS) to provide supplementary meteorologicalinstrumentation at the site. The staff established communications with and utilized the forecasting services

Samples of grass from the area surrounding the plant and the Susquehanna River are shipped out by the EPA to be checked forradioactive contamination.

22studies conducted in other upstream and downstream
areas of the river. Thermal and chemical discharges
during and following the accident did not exceed theeffluent limitation established to protect the aquatic
environment. Although several million gallons oftreated industrial waste effluents were released into
the river, these discharges were not of unusual volume
compared with normal operation and were a verysmall portion of the seasonally high spring river flows.
The extent and relative location of the effluent plumewere defined and the fish species known to have been
under its immediate influence were iden~tified-including rough, forage, and predator/sport
fishery species. Impacts to benthic invertebrates or
fishes were not detected. No unusual conditions of fish
disease or mortality were noted in the river following
the accident. The normal spring increases in abundance and species-composition of riverine fauna occurred, as did the onset of the fish spawning season in
April with peaks of ichthyoplankton abundance in
May and June.
Nevertheless, post-accident recreational fishing in
the Three Mile Island vicinity underwent significant
departures from historical trends. Fishing activity
appeared to shift away from the Susquehanna River
waters near the nuclear station to other areas,
especially downstream. Anglers returned greater proportions of their catches than during any comparable
period within the previous five years. This was mostnotable during April when anglers fishing near theplant returned an unprecedented 100 percent of theircatches. Thus, in the waters receiving station effluentsduring the month following the accident, the liquidradiological pathway leading to man via fin fish consumption could have been absent entirely. With thepassage of time following the accident, the normal
pattern of recreational fishing was approached. The
investigation defined several generic aspects of the accident and lessons applicable to other facilities: thetime of the accident with respect to the biologicalseason, and to the ability to detect an impact; data
availability and data needs for adequate monitoring;and the application of the non-radiological findings
for radiological assessment. This investigation is
described in an NRC report: "The Non-RadiologicalConsequences to the Aquatic Biota and Fisheries of theSusquehanna River from the 1979 Accident at Three
Mile Island Nuclear Station" (NUREG-0596).

Thermoluminescent dosimeter (enlarged at right), used by the
NRC to measure the amount of airborne radiation delivered. to a
specific place, is shown mounted on a utility pole near a school in
Middletown, Pa. Similar devices were installed both on the TMI
site and at various locations around the plant up to 15 miles away

by NRC, the Environmental Protection Agency, Metropolitan
Edison Co., and an independent contractor. In addition to making
independent evaulations, each group sent the data collected toboth the NRC and EPA for analysis.

TMI RECOVERY OPERATIONS
Following the accident of March 28, a substantial
effort was mounted to provide technical assistance,
regulatory guidance and review of the licensee's operations procedures and system addition and modification
activities. A team began to form with the arrival of the
Office of Inspection and Enforcement Region I inspectors shortly after the accident and continued to expand
with the arrival of the first contingent from the Office

====================================================23
of Nuclear Reactor Regulation (NRR) on March 29
and additional inspectors from all five regional offices.
On March 30, the Director of NRR and additional
NRR staff arrived on the site to assist in the recovery
operation. A Public Affairs Office was also establishedin Middletown, PA, and staffed on a 24-hour-per-day
basis to handle the flow of information to the public
and the media.
NRR staff analysts in many of the major disciplines
were brought to TMI to provide needed technicalresources. The specific activities engaged in by thestaff can be broken down into four major areas:(1) A review was initiated of the system modifications and system additions (proposed by thelicensee, the industry review group, or theNRC) as contingency measures to mitigate theconsequences of the accident and to provide
assurance for continued safe shutdown andlong-term safe shutdown.
(2) Substantial effort was given to the review of allprocedures, both emergency and normal operation and maintenance, which were necessary to
post-accident activities. In many cases, becauseof changes in the use of normal systems and the
addition of new ones, new operating procedures
were necessary. Further, the facility license andtechnical specifications, which defined thelimits for operating parameters and surveillance
requirements, were no longer fully applicable tothe post-accident facility, though existing facility procedures provided a mechanism forestablishing specific operability limits andsurveillance requirements. It was necessary,from a regulatory point of view, to have NRC'sreview and approval of any new procedures thatmight be in conflict with the pre-accident
license.
(3) NRR provided close and continuous monitoringof the operations in progress to assure thatsystem parameters stayed within expected limits
and to provide prediction of future system performance and the capability of plant systems tomaintain safe conditions.
(4) Lastly, substantial NRR effort was committed
to providing consultation, review and analysisof the ongoing radwaste, cleanup, and healthphysics activities. The accident generated asignificant amount of contaminated water
which, in turn, contaminated substantial portions of the facility and its systems. This made itdifficult to have normal access to systems important to safety and also constituted a threat offurther fission product release and occupational
exposure. In addition, the radiological makeupof the contamination was different from that
normally encountered in operating reactors, interms of its airborne intensity as well as its ratio

of beta and gamma activity. It was therefore an
important concern-particularly in view of theintensive work activity needed to continue safe
operation-that operator exposures be maintained within acceptable limits and the environment protected from undue radiological effluents.
Examples of the system review activity undertakenby the NRR on-site staff were design reviews andevaluations of the following sytems:(a) Supplementary diesel generators(b) Supplementary filtration systems(c) Long-term cooling systems(d) Alternative decay heat removal system(e) Pressure volume control system(f) Tank farm for storage of radioactive liquids(g) EPICOR-II system for processing of contaminated liquids(h) Many monitor modifications in existing systems
which allowed operability in the post-accident
environment.
Besides the systems reviews, approximately 250 procedures were reviewed and approved by the on-site
staff. This activity was particularly important in thefirst two months following the accident because aserious shortage of personnel familiar with the facility
existed; the NRC review constituted not only aregulatory approval of the intended operation, butalso served as a quality assurance check on adequacy
and operability. The review of procedures is continuing as the licensee rewrites emergency and operatingprocedures to reflect the changing status of the facility. It is anticipated that such procedure review will benecessary until a new set of facility technical specifications, which reflect the post-accident facility configuration, is implemented.
A substantial amount of staff effort was expended
on the review and approval of the EPICOR-II operation, intended for use in decontaminating the 380,000
gallons of intermediate-level contaminated water heldin the auxiliary building tanks and in the tank farmconstructed following the accident. EPICOR·II was
designed and constructed following the accident
because it was clear that storage of water would be asignificant problem and could not be accommodatedwith existing facility equipment. EPICOR-II is athree-stage demineralization system, constructed in anexisting on-site building. EPICOR-II was provided
with sufficient shielding and remote-handlingcapability to accommodate the processing of radioactive water up to a level of about 100-microcuries-permilliliter. When facility operation was near, court actions were initiated to prevent operation of EPICORII or disposal of the decontaminated water. In
response to these actions, the Commission directedthat an environmental assessment for the use of

24

The "EPICOR-II" system being used to decontaminate some
380,000 gallons of intermediate-level radioactive water held in theauxiliary building tanks at the TMI.2 site is shown above. It consists of three process vessels (steel liners) shielded by four-inch leadenclosures located in the chemical cleaning building. Each vessel
contains ion-exchan,ge resin. The vessel at the top of the photoat the left is the system prefilter/demineralizer, the center vessel is acation ion-exchanger, and the third vessel is a mixed-bed polishingion-exchanger. Each is fitted with three quick-disconnect hoses: aliquid waste influent line, a processed waste effluent line, and avent line with attached overflow hose. Vented air from each vessel

passes through a special filter and charcoal adsorber. "Spent" ionexchange resin liners containing radioactive material removed fromthe water are transferred by crane to cells (shown at top right)
which are housed in modular concrete storage structures (above).
The cells are concrete-shielded, galvanized corrugated steel
cylinders seven feet in diameter and 13 feet hi,gh. The storage
module shown under construction has 4-foot thick walls and will
be 57 feet wide and 91 feet long. The modules, each holding about
60 storage cells, will be built on an as-needed basis. Shipment of
the radioactive liners away from the site will depend on approvalof a disposal facility and availability of shipping casks.

EPICOR-II be prepared, followed by the environmental assessment for the alternatives of disposal of decontaminated water. Both of these environmental
assessments would be provided to the public for comment before any actions would be initiated. Environmental assessment for the use of EPICOR-Il in
the decontamination of the intermediate level of contaminated water in the auxiliary building was
prepared and sent out for public comment on August
14, 1979. The assessment evaluated various alternatives to the proposed cleanup and concluded thatthe use of the already constructed system was the bestalternative, and that the processing of water would
constitute a negligible environmental impact.

BULLETINS AND ORDERSTASK FORCE

Based on these evaluations, the Commission, on October 16, 1979, issued a Memorandum and Orderdirecting the use of EPICOR-II.

The accident at TMI-2 involved a feedwater transient coupled with a small break in the reactor system(the open relief valve). Because of the severity of theensuing events and the potential generic implicationsof the accident for other operating reactors, the NRC
staff initiated prompt action to: (1) assure that otherreactor licensees, particularly those with plants similar
in design to TMI-2, took the necessary action to
substantially reduce the likelihood for TMI-2 type
events, and (2) start comprehensive investigations into
the potential generic implications of this accident onother operating reactors.
The Bulletins & Orders Task Force was establishedwithin the Office of Nuclear Reactor Regulation(NRR) in early May 1979. This task force was responsi-

===================================================25
ble for reviewing and directing the TMI-2 related staff
activities regarding loss-of-feedwater transients andsmall break loss-of-coolant accidents for all operating
reactors. The task force concentrated its efforts in theareas of: assessments of auxiliary feedwater system
reliability; review of the analytical predictions ofplant performance for both feedwater and small
LOCA-induced transients; evaluations of genericoperating guidelines; the review of emergency plant
operating procedures; and the review of operator
training.
The task force worked with operating plant licenses,
and, for the review of generic items, with owners'
groups for plants of each nuclear steam supply vendor(Babcock and Wilcox, Westinghouse, Combustion
Engineering, and General Electric) and with the indilvidual vendors. Initial priority was placed on plantsof the Babcock and Wilcox (B&W) design, but asshort-term actions on these plants were completed,priority was shifted to other pressurized water reactor(PWR) plants, i.e., those manufactured byWestinghouse and Combustion Engineering. Activities related to boiling water reactors, a significantly
different light water reactor type manufactured by theGeneral Electric Company, were pursued as a third
priority.
The task force, which was composed of approximately 30 technical professionals of widely varying
disciplines and areas of expertise, evaluated licensees'
responses to NRC Bulletins; the issuance and subsequent lifting of Orders to the B&W operating reactors;
system reliability and predicted plant performance foreach of the reactor vendors, with regard to feedwatertransients and small break loss-of-coolant accidents;
and related follow-on activities.

Bulletins
The preliminary review of the accident chronology
identified several events that occurred during the accident and contributed significantly to its severity. As aresult, all holders of operating licenses were subsequently instructed to take a number of immediate
actions to avoid repetition of these errors. These instructions were specified in a series of bulletins issued
by the NRC's Office of Inspection and Enforcement(IE).
The initial bulletins defined actions by operating
plants using the B&W reactor system, but as staff
evaluations determined that additional actions werenecessary, these bulletins were expanded, clarified,and issued to all operating plants for action. For example, as a result of staff evaluations, holders ofoperating licenses for B&W designed reactors were instructed by IE Bulletins to take further actions, including immediate changes to decrease the reactor
high pressure trip point and increase the pressurizer
pilot-operated relief valve settings. A chronology ofbulletins issued by IE is shown below.

The task force directed the evaluations of eachlicensee's response to the IE Bulletins. This process involved an inter-office review group, which included
representatives from IE and from the NRR Division of
Operating Reactors. When it was concluded that alicensee had understood and had provided an acceptable response to the bulletins, the bulletin review wascompleted and the evaluation issued as a staff report.
The prompt action taken by licensees in response totfie IE Bulletins was considered an important contributor to the assurance of continued safe plant
operation. The bulletins and related evaluations also
provided substantial input to other staff activities,
mch as those associated with the generic study efforts
and the Lessons Learned Task Force (see below).
Thus, many of the subjects addressed by the bulletinswere studied in greater depth through other staff activities and studies. Further, the bulletins and theassociated responses were used as a basis for IE inspection activities and for auditing of reactor operator
training.

Orders on Babcock and Wilcox Plants
Soon after the TMI-2 accident, the NRC staff begana reevaluation of the design features of B&W reactors
to determine whether additional safety corrections or
improvements were necessary. This evaluation involved numerous meetings with the vendor and the affected licensees.
The conclusion of these preliminary staff studies was
documented in an April 25, 1979 status report to theCommission. It was found that B&W designed reactors appeared to be unusually sensitive to certain transient conditions originating in the secondary system.
The features of the B&W plants that contributed tothis sensitivity were: (1) design of the steam generators
which operate with relatively small liquid volumes inthe secondary side; (2) lack of direct initiation of reactor trip upon the occurrence of off-normal conditionsin the feedwater system; (3) reliance on an integrated
control system (ICS) to automatically regulate feedwater flow; (4) actuation before reactor trip of a pilotoperated relief valve on the primary system pressurizer(which, if the valve sticks open, can aggravate theevent); and (5) a low steam generator elevation(relative to the reactor vessel) which provides a smaller
driving head for natural circulation (except for theDavis-Besse plant).
Because of these features, B&W design relies morethan other PWR designs on the reliability and performance characteristics of the auxiliary feedwater
system, the integrated control system, and theemergency core cooling system (ECCS) performance
to recover from certain anticipated transients, such as
loss of off-site power and loss of normal feedwater.
This, in turn, can require greater operator knowledge
and skill to safely manage the plant controls during

26

such anticipated transients. As a result of the work
supporting the April 25, 1979 report, the NRC staff
concluded that certain other short-term design andprocedural changes at operating B&W facilities werenecessary in order to assure adequate protection to
public health and safety.
After a series of discussions between the NRC staff
and licensees of operating B&W plants, the licensees
agreed to shut down these plants and keep them shut
down until the actions identified to the Commission in
the April 25, 1979 report could be completed. Thisagreement was confirmed by a Commission Order toeach licensee (see "Actions Directed by Orders,"
below). Authorization to resume operation was issued
in the period late May through early July, as individual plants satisfactorily completed the short-term
actions and the NRC staff completed an on-siteverification of the plant's readiness to resume operation. In addition to the modifIcations to be implemented promptly, each licensee also proposed tocarry out certain additional long-term modifications
to further enhance the capability and reliability of theplant systems to respond to transient events (see"Longer Term Actions," below).
Since some of the long-term modifications involvethe design, procurement, and qualification of safetygrade hardware, not all of the actions of the long~term
portion of the Orders were completed in 1979. Staff involvement will continue to assure that licensees complete each long-term action of the Order "as promptly
as practicable" and that the Orders are closed out by aprompt staff acceptance review.

Specific Plant and Generic Studies
For 13&W operating reactors, an initial staff study
has been completed and published in a staff report(NUREG-0560). This study considered the particular
design features and operational history of B&W
operating plants in light of the TMI-2 accident andrelated current licensing requirements. As a result of
this study, a number of findings and recommendations
resulted which are now being pursued.
Generally, the activities involving the B&W reactors are reflected in the actions specified in the Orders.
Consequently, as noted earlier, a number of specificactions have been specified in the areas of transientand small break analyses, upgrading of auxiliary feedwater reliability and performance, procedures for
operator action, and operator training.
Similar studies are now well underway for the Westinghouse and Combustion Engineering operating

plants. These studies focus specifically on thepredicted plant performance under different accident
scenarios involving small break loss-of-coolant eventand feedwater transients. Based upon analytically
predicted system behavior, recommended guidelines
for emergency operating procedures were developed
and reviewed in the study. In addition, these studies
include engineering assessments of the reliability of individual plant auxiliary feed water systems and identification of dominant failure contributors and recommendations for corrective action. A similar study ofthe operating boiling water reactors is also in progress,
but is at an earlier stage.
As the above studies developed firm conclusions andrecommendations, implementing action was initiated.
For example, the results of the Westinghouse andCombustion Engineering auxiliary feedwater system
reliability assessments concluded that certain improvements were necessary. Individual plant licensees
were then requested by letter to initiate corrective action or to propose design solutions for NRC staff
review. Additional instructions were to be issued tolicensees upon completion of other aspects of thesereports.

Follow-On and Interfacing Activities
It was planned that the task force would terminate
its activities in late 1979, and therefore some of its activities were transferred prior to completion. Consequently, the task force concentrated on lead plants andestablished review guidelines and acceptance criteriathat could be implemented by other NRR organizational elements.

As a result of the work performed in modeling small
break and feedwater transients, longer range efforts
were identified dealing with the procedures andsystems available for core cooling under certain accident conditions, and with confirming analyticalmodels through experiment or research programs.
For example, plans are being implemented to conduct
some small break loss-of-coolant tests at the Semiscaleand LOFT facilities to obtain a better understandingof small break phenomena and to use the results toverify calculational techniques (see Chapter 11).
Other recommendations in this regard are expected toresult from the task force activities.
As noted previously, the task force concentrated on
the immediate and near term actions necessary toassure the safe operation . of operating pl~nts.
However, based on actions already completed. anumber of items have been identified which warrant
careful additional study. These actions haye been andare continuing to be, documented for detailed assessment within the NRR organization.

27

IE BULLETINS ISSUED: APRIL-JULY 1979
Bulletin

Date Issued

Issued to

79-05
79-05A
79-06
79-06A
79-06B
79-08
79-06A (Rev. 1)
79-05B
79-05C
79-06C

April 1, 1979
April 5, 1979
April 11, 1979
April 14, 1979
April 14, 1979
April 14, 1979
April 18, 1979
April 21, 1979
July 26, 1979
July 26, 1979

B&W plants
B&W plants
Wand CE plants
Wand CE plants
CE plants
BWR plants
W plants
B&W plants
B&W plants
Wand CE plants

Actions Directed by NRC Orders(for immediate implementation)(1) Reviewing and upgrading, as appropriate, auxiliary feed water reliability and performance.
(2) Implement operating procedures for initiating
and controlling feedwater independent of ICS.
(3) Implement hard-wired control grade reactor
trip on loss of main feedwater and/ or turbine
trip.
(4) Complete analyses for potential small breaks
and implement appropriate instructions for
operator action.
(5) Provide at least one senior reactor operator,
having TMI-2 training on B&W simulator, incontrol room.

Longer-Term Actions Required by Orders(1) Continue to review and upgrade performance ofauxiliary feedwater system.
(2) Submit a failure mode and effects analysis of theintegrated control system to the NRC.
(3) Improve the quality of the reactor trip following
loss of main feed water and/ or turbine trip byupgrading to safety-grade design.
(4) Give continued attention to transient analysis
and procedures for management of small
breaks.

TMI-2 LESSONS LEARNEDTASK FORCEIn May 1979 an interdisciplinary team of engineersfrom the NRC Offices of Nuclear Reactor Regulation,
Nuclear Regulatory Research, Inspection and Enforcement, and Standards Development began work on theidentification and evaluation of safety concerns
originating from the TMI-2 accident that required
licensing actions. This team, the TMI-2 Lessons
Learned Task Force, concentrated on issues separate
from those specified in IE Bulletins and Commission
Orders issued to operating plants early after the accident. The areas of interest to the Lessons Learned Task
Force were applicable not only to operating plants but
also to pending operating license (OL) and construction permit (CP) applications.
The task force was charged to review and evaluate
investigative information, Commissioners' recommendations, ACRS recommendations, staff recommendations from NUREG-0560 ("Staff Report on the Generic
Assessment of Feedwater Transients in Pressurized
Water Reactors Designed by the Babcock & Wilcox
Company"), and recommendations from outside the
NRC. In addition, the task force was charged to identify, analyze and recommend changes to licensing requirements and the licensing process for nuclear
power plants based on the lessons learned. The scopeof the task force assignment covered the following
general technical areas:(1) Reactor operations, including operator training
and licensing.
(2) Licensee technical qualifications.
(3) Reactor transient and accident analysis.
(4) Licensing requirements for safety and processequipment, instrumentation, and controls.
(5) On-site emergency preparations and procedures.
(6) NRR accident response role, capability andmanagement.
(7) Feedback, evaluation, and utilization of reactor
operating experience.

Two Phases of Work
The work of the task force proceeded in two phases.
The first was the development of recommendations for
short-term actions which, when combined with the requirements associated with implementation of the IE
Bulletins on TMI-2-including the generic status
reports issued by the task force and certain otherchanges. in emergency preparations by licensees andoperator training and licensing requirements-would
ensure the safety of plants already licensed to operate
and those to be licensed for operation in the near
future. The first phase culminated with issuance in July 1979 of a report entitled "TMI-2 Lessons Learned
Task Force: Status Report and Short-Term Recommendations" (NUREG-0578). The implementation of
23 short-term licensing requirements was directed for
operating reactors by the Director of NRR in
September 1979 based on a favorable ACRS review
received in August.

28(1) Emergency Power. For PWRs (pressurizedwater reactors), provide emergency power forthe minimum number of pressurizer heaters required to maintain natural circulation conditions in the event of loss of off-site power, forpower-operated relief valves and associatedblock valves, and for pressurizer level instrument channels.
(2) Valve Tests. For BWR (boiling water reactors)
and PWR relief and safety valves, perform fullscale performance verification tests.
(3) Valve Position Indication. Provide direct position indication for PWR and BWR poweroperated relief valves and safety valves.

Metropolitan Edison staff members work in a room adjacent to the
TMI control room to coordinate communication between the plant
and local officials such as State police and fire departments.

In the second phase of its work, the task force considered more fundamental questions in the design andoperation of nuclear power plants and in the licensing
process. The issues were grouped in four general categories: general safety criteria, system design requirements, nuclear power plant operations, andnuclear power plant licensing. A report entitled"TMI-2 Lessons Learned Task Force: Final Recommendations" (NUREG-0585) was issued in October
1979 to complete this phase.
The completion of these reports terminated the formal activities of the Lessons Learned Task Force, andits members returned to other duties. Two small
groups among them, however, remained intact tomake up the nuclei of interdisciplinary review teams
which will see to the implementation of task force
recommendations for new operating licenses and for
operating reactors.

Short-Term Recommendations
The decisionmaking process followed by the task
force in determining which safety issues required
short-term licensing action versus those that could be
deferred for further evaluation by the task force orothers was based on engineering evaluation andqualitative professional judgment of the safetysignificance of the various issues. In this regard, thetask force selected items for "short-term action" iftheir implementation would provide substantial, additional protection required for the public health andsafety. The task force recommendations presented in
NUREG-0578 comprised 23 specific requirements.
Each of these is discussed in detail in NUREG-0578,
and a proposed two-stage implementation schedule isincluded as an appendix to that report. The 23 recommendations are briefly stated below.

(4) Instrumentation jor Inadequate Core Cooling.
Perform analyses and implement procedures
and training for prompt recognition of low reactor coolant level and inadequate core coolingusing existing or modified instrumentation;
analyze and describe instrumentation for detection of low reactor vessel water level.
(5) Containment Isolation Signals. Provide containment isolation on diverse signals, review isolation provisions for non-essential systems andrevise as necessary, and modify containment
isolation designs as necessary to eliminate thepotential for inadvertent reopening upon resetof the isolation signals.
(6) Recombiner and Purge Penetrations. For plantsthat have external hydrogen recombiners orpurge systems, provide dedicated penetrations
and isolation systems that meet the redundancy
and single failure requirements of the Commission regulations.
(7) Inerting BWR Containments. Provide inerting
for all Mark I and Mark II BWR containments.
(Rulemaking required.)(8) Hydrogen Recombiner Capability. Provide thecapability to add, within a few days after an accident, a hydrogen recombiner system for postaccident hydrogen. (Minority view; rulemaking
required.)(9) Systems Leakage. Perform leakage rate tests on
systems outside containment that process
primary coolant and could contain high level
radioactive materials. Develop and implement
periodic testing and preventive maintenanceprograms.
(10) Shielding Review. Perform a design review ofthe shielding of systems processing primarycoolant outside containment and assure that access to vital areas will not be unduly impaired
due to radiation from these systems.
(11) Automatic Initiation oj the Auxiliary Feedwater
System. Provide means for automatic initiation

====================================================29of all auxiliary feedwater systems; manual
capability to initiate the auxiliary feedwater
system from the control room must be retained.
(12) Auxiliary Feedwater Flow Indication. Provide

indication in the control room of auxiliary feedwater flow for each steam generator.
(13) Post-Accident Sampling. Review and upgrade
the capability to obtain and analyze samples
from the reactor coolant system and containment atmosphere under high radioactivity conditions.
(14) High-Level Radiation Monitors. Provide highrange radiation monitors for noble gases inplant effluent lines and a high-range radiation
monitor in the containment. Provide instrumentation capable of measuring and identifying
radioiodine and particulate radioactive effluents in effluent lines under accident conditions.

(15) Improved In-Plant Iodine Instrumentation.
Provide instrumentation for accurately determining in-plant airborne radioactive concentrations to minimize the need for unnecessary useof respiratory protection equipment.
(16) Analysis of Transients and Accidents. Provide
the analysis, emergency procedures, and training to improve operator performance during asmall break loss-of-coolant accident, to assurethat the reactor operator can recognize and respond to conditions of inadequate core cooling,
and to improve operator performance duringtransients and accidents, including events thatare caused or worsened by inappropriate operator actions.
Shift
Supervisor Responsibilities. Promptly is.sue(17)
an operations policy directive that emphasIzes
the duties, responsibilities, and authority andlines of command of the control room operators,
the shift supervisor, and the person responsible
for reactor operations command in the control
room.

(20) Control Room Access. Revise emergency procedures as necessary to assure that access to thecontrol room under normal and accident conditions is limited to those persons necessary to thesafe command and control of operations.
(21) On-site Technical Support Center. Provide an

on-site technical support center, separate fromthe control room, for use by plant management,
and technical and engineering support personnel in an emergency. This center shall be used
for assessment of plant status, support of thecontrol room command and control function,and in conjunction with implemen'tation of onsite and off-site emergency plans. Communications links shall be established and the center
shall be equipped as necessary for emergency
engineering support activities.
(22) On-site Operational Support Center. Establish
and maintain an on-site operational support
center to which shift support personnel (e. g. ,auxiliary operators and technicians) other thanthose required and allowed in the control room
report for further orders and assignment during
an emergency.
(23) Loss of Safety Function. Require that a reactor
be shut down if human errors lead to a complete
loss of safety function (e. g., loss of emergency
feedwater, high pressure ECCS, low pressure(18) Shift Technical Advisor. Provide a shift technical advisor at each nuclear power plant who
has a bachelor's degree or equivalent in ascience or engineering discipline and withspecific training in the plant response to offnormal events and in accident analysis of theplant.
(19) Shift and Relief Turnover Procedures. Review
and revise plant procedures as necessary toassure that a shift turnover checklist is provided
and required to be completed and signed by the
oncoming and offgoing individuals responsible
for command of operations in the control room.

Portable communication units provided by the U.S. Forest Service
were used to communicate between the TMI control room andvarious staff activities at the site. One such unit was manned on a
24-hour basis while periodic checks were made with the control
room to record the status of the reactor.

30

ECCS, containment, emergency power or otherprescribed safety function), and allow the facility to return to power only after a public meeting
and NRC approval of the remedial changes proposed by the licensee. (Rulemaking required.)
After considering comments on NUREG-0578 byvarious NRC offices, the Advisory Committee on
Reactor Safeguards (ACRS), the industry and the
public, the Director of Nuclear Reactor Regulation,with the approval of the Commission, added four requirements as follows:(1) Containment Pressure Indication (ACRS). Provide wide-range continuous indication of containment pressure in the control room.

(2) Containment Hydrogen Indication (ACRS).
Provide continuous indication in the control
room of hydrogen concentration in the containment atmosphere.
(3) Containment Water Level Indication (ACRS).
Provide continuous indication in the control
room of containment water level.
(4) Reactor Coolant System Vents. To provide
means for removing noncondensible gases, install reactor coolant system and reactor vessel
head high point vents remotely operated fromthe control room.

Implementing Short. . Term
Recommendations
The Commission directed that the staff proceed assoon as possible with implementation of all of theshort-term recommendations, except those which weremodified as set forth below, on the two-stage,
16-month schedule recommended by the task force.
In view of ACRS comments, the Director of Nuclear
Reactor Regulation decided to delay any rulemaking
action concerning inerting of BWR Mark I and II containments and provisions of hydrogen recombiner
capability at operating plants until the final report of
the task force had been issued. Final resolution of thesematters is discussed in the section below covering thelong-term recommendations of the Lessons Learned
Task Force.
With respect to the recommendation to add a Shift
Technical Advisor at each plant, the ACRS endorsed
the concept but suggested that flexibility be maintained in implementation so that the objective could
be reached through innovative approaches by individual licensees. For guidance, the task force prepared a statement of functional characteristics for the
Shift Technical Advisor to be used in evaluating alternatives proposed by licensees.
The recommendation to review limiting conditionsof operation to incorporate mandatory shutdown if
human error causes loss of a safety function stimulated

much interest inside and outside the staff. The Officeof Standards Development has prepared a paper proposing such a new rule, but setting forth alternatives
for achieving the same objectives as the task force
recommendation.
On September 13, 1979, letters were sent to all
operating nuclear power plants advising them thatthey should proceed with implementation of therecommendations of the Lessons Learned Task Forceand the additional items resulting from ACRS comments and review by the Director of Nuclear Reactor
Regulation. During the week of September 24, 1979,
regional briefings were held to apprise reactor ownersof these requirements. These meetings were followed
by a 3-day series of meetings at NRC headquarters in
Bethesda, Md. on some of the specific short-term requirements. Letters were also sent to applicants forconstruction permits and operating licenses instructing
them to implement the short-term lessons learned.
All of the short-term "Category A" requirements
deriving from conclusions of the Lessons Learned Task
Force were conveyed to licensees of operating reactors
by the end of 1979. It was expected that about twothirds of these licensees would have met the Category
A requirements by the end of January 1980, and therest by May at the latest.
The approach adopted by NRC staff in seeking swift
implementation of the short-term requirements allowed licensees to fulfill those requirements prior to NRC
staff review. The approach necessitated a carefulclarification of each requirement, and this was provided by means of regional as well as topical meetings andnumerous discussions among NRC staff, the vendororiented owners' groups, and licensees.
The small number of action items that were notcompleted by the deadlines prescribed by NRC mainly
involved problems of equipment availability. Someslippage is also permitted where it can be demonstrated that a severe impact on regional power supply
would otherwise result.

Long Term Recommendations
In contrast to the short-term recommendations in
NUREG-0578, which were of a more narrow, specific,
and urgent nature, the final report of the task force(NUREG-0585) dealt with safety questions of a more
fundamental policy nature regarding nuclear plant
operations and design and the regulatory process.
To stimulate discussion and speed the deliberative
process, the task force developed a number of specific,
final recommendations toward accomplishing the. policy objectives and safety goals described in thereport. The task force considered the modifications it
outlined to be of fundamental importance to nuclear
safety, and urged that immediate steps be taken tocomplete the deliberative process and initiate implementation of the recommendations.

=====================================================31
Although the accident at Three Mile Island
stemmed from many sources, the most importantlessons learned fall in a general area the task force
chose to call operational safety. This general area in~cluded the topics of human factors engineering; qualification and training of operations personnel; integration of the human element in the design, operation,and regulation of system safety; and quality assuranceof operations. Specifically, the primary deficiency inreactor safety technology identified by the task force's
review of the accident was the inadequate attention
that had been paid by all levels and all segments of thetechnology to the human element and its fundamentalrole in both the prevention of accidents and the
response to accidents. Thus, the policy recommendations and specific ideas in NUREG-0585 forstimulating and accomplishing change concentrated
heavily on operations reliability and the associateddesign and licensing review measures that support oraugment operations reliability.
The task force also devoted considerable attention to
the basic mission of reactor regulation after Three Mile
Island. It was not alone in these efforts; many people
called for a clearer articulation of NRC's role and mission after March 28, 1979. The task force found thatprescriptive and narrow licensing requirements only
add to the quiltwork of regulatory practice and dolittle to directly address the nation's heightened concern for the safety of nuclear power plants. The task
force called for the development of an articulate andwidely noticed national nuclear safety policy withwhich to bind together the narrow and highly technical licensing requirements. Although the NRC andthe President's Commission alluded to a moredefinitive safety policy by taking actions that in effectsay, "no more Three Mile Islands," the task force
urged that the feasibility and the adequacy of such a
policy be critically examined and an opportunity provided for thorough and widespread public input.
More than 30 recommendations in 13 different areas
were made by the TMI-2 Lessons Learned Task Force.
In its review of these recommendations, the ACRS
supported them in all 13 areas, offered advice on
details of implementation and criteria employed insome of those areas, and added comments and recommendations on areas not addressed in the task force
reports. Final recommendations of the task force andof the ACRS were being factored into the developmentof the NRC Action Plan for TMI-2 matters, which wasin preparation at the end of 1979.

Activity in the trailer office of NRC's Office of Inspection and Enforcement shortly after the accident, where personnel kept track of
the environmental monitoring of radioactivity. At far right is~ichard H. Vollm~r of N~C's Office of Nuclear Reactor RegulalIon, who was deSignated 10 June to direct NRC's support activities
related to recovery and cleanup operations at the TMI site.

32"Herman," a mobile manipulator borrowed from the Oak Ridge
National Laboratory, proved too awkward for use at TMI-2. It
was hoped that the robot could retrieve samples of radioactive
water in the No.1 Auxiliary Building, thereby reducing exposure
to workers. The idea was abandoned when testing showed Herman's lack of pressure sensitivity presented the risk of flasks of contaminated water being dropped or crushed.

Inspection and Enforcement Lessons
The NRC Office of Inspection and Enforcement(IE) also undertook an intensive investigation of the
TMI accident but limited the scope of its inquiry to
two sharply defined aspects of it: (1) the operational
activities of the licensee from before the initiating
event, about 4 a.m. on March 28 up to about 8 p.m.
that evening, when primary coolant flow was reestablished by the starting of the reactor coolant pump; and(2) steps taken by the licensee to control the release ofradioactive material to off-site environs and to implement its emergency plan, from the initiating event until midnight on March 30. These periods were selected
for scrutiny because, in the judgment of IE, theyencompassed those licensee actions which mostsignificantly affected the course of the accident and its
consequences.
In its report on this investigation, issued August 3,
1979, IE confirmed that the collective radiation doseto the general public resulting from the TMI accident
constituted-as reported by the Ad Hoc Dose Assessment Group (made up of various Federal agencies) in
its May 10, 1979 report-minimal risk to the health ofthe off-site population. At the same time, IE reported
several inadequacies in the licensee's radiation protection activities inside the plant, as well as in the measuring of off-site radiation levels. These flaws, however, were not such as would cast doubt on or call foralterations in the conclusions of the ad hoc group.

The IE investigation also substantiated earlier con·clusions regarding the underlying causes of the TMI
accident and the factors that contributed to its severity. The six distinct areas of deficiency earlier identified
as causative or complicating elements and confirmed
by IE comprised equipment performance; licensee
analysis of past transients and accidents; operator
training and performance; equipment and systems
design; the transmission of information (especially inthe early phase of the accident); and the implementation of emergency planning. But what the IE report
called "perhaps the most disturbing result" of the investigation was "confirmation of earlier conclusions
that the Three Mile Island Unit 2 accident could havebeen prevented, in spite of the inadequacies" cited.
The design, equipment, analyses, and procedures inplace and in effect at TMI were, IE concluded, sufficient "to have prevented the serious consequences of
the accident" if they had been allowed to function or
had been adhered to as intended. For example, had
the TMI operators permitted the ECCS to have its full
effect, the damage to the core would most likely havebeen prevented (other examples were adduced in thereport where a right action taken or a wrong one
avoided could have significantly mitigated the consequences of the accident).
On the other hand, the IE report concedes, had certain equipment been designed differently it too could
have prevented or diminished the effects of the accident. The investigation made it "difficult to fault only
the actions of the operating staff." An undue preoccupation with the hazards of overfilling the reactor
coolant system (that it was to be avoided "at almost
any cost") was also evident in the decisions and actionsof the operators, leading them to ignore prescribed
procedures and to fail to respond to indications thatthe core was not properly cooled. Retraining of all
licensed operators has now been required by NRC as
well as an upgrading of procedures.
Causes and Contributing Factors. Soon after theshift came on at TMI-2 at midnight of March 27, 1979,
the shift foreman and two auxiliary operators were
engaged in transferring resin from a "condensate
polisher" tank to a "resin regeneration" tank, on thesecondary side of the plant. The chore was a carryoverfrom the previous shift and was one with which plant
personnel had encountered some difficulty. The staff
thought the _problem was a resin blockage in thetransfer line and the foreman and auxiliary operators
were trying to clear it. The IE report concluded that,"probably as a result of their efforts to clear the line,"
the plant underwent a total loss of feedwater flow, initiated by a loss of condensate flow and bringing about
an almost simultaneous shutdown of the main turbine,
at 37 seconds after 4 a.m., on March 28.
Ensuing events were found to be as described earlier
in this chapter with certain noteworthy additions andconclusions. Among these was the finding that, about

33

six minutes after the start of the accident, thepressurizer was completely filled with water and thereactor coolant system was, in fact, "solid," the condi~tion which the control room crew strived to avoid
throughout the crucial early hours of the accident byactions which delayed cooling of the core and com~pounded the consequences of the event. The IE report
also indicated that "substantial fractions of the core
were uncovered" by about 6:30 a.m. on March 28,
although the fact went unrecognized by the operatorsand officials on the scene, and the high temperature
readings in the core and the loops were considered too
high to be realistic. The report also found that theoperators interpreted the failure of the core flood tanks
to inject a substantial portion of their volume into thereactor coolant system to be an indication that the core
was covered, even though these tanks cannot be used
for that purpose and are designed to supply water inthe event of a large loss~of-coolant accident, which
was not happening. With respect to the hydrogen explosion in the containment, the report observed thatthe release of this noncondensible gas from the reactor
coolant system may have contributed to the later apparent success of the staff in collapsing the voids in the"A" loop of the reactor. That appearance of success inestablishing natural circulation, despite the continued
high temperatures in portions of the system, led theoperators to believe that they had attained areasonably stable condition by early afternoon ofMarch 28.
Specific actions cited by the IE report as bringing
about the extensive core damage that took place inCluded: throttling the high pressure injection (ECCS)
to a minimum during the first three and one-half hoursof the accident; operating the reactor coolant pumps
at pressures below procedural requirements (which led
to greater loss of coolant through the stuck-openpressurizer relief valve); failure to isolate the relief
valve after pressure continued to fall in the reactor
coolant system, the drain tank disc had blown, and thes~mp pump operation all indicated that a large
discharge of water from the system and the building
was taking place; and failure to establish the conditions necessary for natural circulation in the system.
The report made note of other licensee actions
which, while they did not directly affect the course of
the accident as it actually unfolded, could have severely impaired the response of safety-related equipment ifthat course had taken another direction. Specifically,
the disabling of the automatic startup features of theemergency diesel generators and the isolating of thecore flood tanks early in the event constituted thesekinds of lapses. The report was also critical of the communications provided during the event by the licensee,
pointing out that persons assigned to furnish information off-site had concurrent duties related to management of the emergency. At the root of this and otherproblems, the report concluded, lay the misconceptionthat even major accidents would be short-term eventsand that plans for mobilizing and communicatingwith off-site technical support over time, as an accident progresses, was not warranted as part of theemergency planning.
Enforcement Action Proposed. On October 25,
1979, the Director of Inspection and Enforcement
notified the licensee for TMI that the IE investigation
had revealed "numerous items of noncompliance"with NRC regulations on the part of the licensee. Six"violations" -the most serious breach of regulatory requirements-were alleged by IE, including serious
weaknesses in the licensee's health physics program;
control of maintenance activities; development andreview of procedures; adherence to prescribed procedures; and audit activities. The licensee was citedfor failure to operate the facility in accordance withthe technical specifications approved and adopted forthat particular plant, and for authorizing asurveillance procedure which placed certain valves ina status which rendered emergency feedwater
unavailable on three separate occasions-including
the last on March 28, when it was needed. Personnel
training was also found insufficient, and record
maintenance and in-house inspections as well.
The licensee was called upon to correct each of these
deficiencies and departures from requirements andwas notified that civil penalties were being proposedin the amount of $155,000, the legal maximum(although an assessment of $725,000 was justified forall violations identified).
Task Force Urges Statutory Mandate on Lead Role.
The IE task force on leassons learned from TMI urgedthat IE be assigned, by statutory mandate, the leadrole in NRC's emergency response in the future. Such arole flows from IE's de facto activity as the "principal
contact with operating licensees," it was argued. It
was also recommended that intra-office training be expanded and tightened surveillance of licensees be
adopted. In the lead role for NRC emergency
response, IE could give assistance to licensees in its
response to an incident, as well as coordination to all
NRC acitivities. It would also undertake training ofother NRC offices regarding emergency preparedness
and the respective responsibilities of those offices.
The task force also recommended that NRC create a
new office to oversee training programs to upgrade thequality of inspectors and operations personnel,
especially in the area of emergency response.
ACRS Comment on IE Findings. In a letter toChairman Hendrie dated November 14, 1979, the Advisory Committee on Reactor Safeguards (ACRS)
registered its view of the IE investigation and conclusions based on that investigation. Taking note of thelimitation in scope of the IE study, the ACRS felt thatthe emphasis put by IE on the licensee's departurefrom technical specifications prior to the accident andfrom approved procedures during it resulted in too lit-

34

A new ~entilation filtration system was installed on top of the auxiliary building of Unit 2 when the NRC determined that the existing system was not functioning satisfactorily after the accident.

This system filters out airborne radioactive iodine and particulates
before air is released to the environment.

tIe consideration of other relevant factors. Examples ofsuch factors taken from other investigations by NRC
and others might be the peculiarities of a nuclear
steam supply system that tended to inhibit recovery
from interruption of normal operation or to confuse
the operators by producing conditions and instrument
readings not anticipated in the written procedures
and, in general, by failing to convey clear, complete
information to those in the control room. The ACRS
concluded that the limited scope of the IE report tended to lead to a catalogue of violations and expressed its
concern that the rationale behind the IE report would
be perceived to be that a licensee's failure to follow accident procedures is automatically a violation. The
ACRS noted that the procedures are prepared by thelicensee and are not approved by NRC (although thelicensee is required by NRC to follow them) and affirmed that such procedures cannot be so detailed as toallow for every accident situation. On the contrary,

the ACRS declared, a deviation from conditions
assumed in the framing of procedures may make itnecessary to depart from those procedures. There is aquestion as to whether an operator who, using his best
judgment, consciously takes an action at variance withprocedures which in themselves may contain confusing
or incorrect guidance is guilty of a violation. If this is
the case, the ACRS affirmed its belief that it is "thewrong approach to protecting the public health andsafety" in an emergency and that an operator, guided
by written procedures, should be allowed to use hisbest judgment to deal with a problem. That judgment
would be subject to post-factum appraisal by responsible parties, but it should not necessarily be deemed anerror or a violation of regulations.
The ACRS found the IE report "less than satisfactory" for these reasons and recommended issuance of aconsolidated report on the findings of the several NRC
task forces investigating the TMI accident.

35

The President's Commission
On April 11, 1979, President Carter issued an executive order (# 12130) creating the President's Commission on the Accident at Three Mile Island andcharging its members to "conduct a comprehensive
study and investigation of the recent accident involving the nuclear power facility on Three Mile Island in
Pennsylvania" and to include in their study the following elements:• A technical assessment of the events and theircauses.
• An analysis of the role of the managing utility.
• An assessment of the emergency preparedness andresponse of the NRC and other Federal, State andlocal authorities.
• An evaluation of the NRC's licensing, inspection,
operation and enforcement procedures as applied
to this facility.
• An assessment of how the public's right to information concerning the events at Three Mile
Island was served and of the steps which should
be taken during similar emergencies to provide
the public with accurate, comprehensible andtimely information.
• Appropriate recommendations based upon theCommission's findings.
The President appointed John G. Kemeny, thePresident of Dartmouth College and former chairmanof the Mathematics Department at that institution, tothe chairmanship of the Commission. Eleven othermembers were appointed, including a State Governor,a resident of Middletown, Pa., a labor union president, an industrialist, the president of a national society, an attorney, and five unversity professors. A fulltime staff was engaged which eventually numberedover 60 persons; more than 30 separate staff reports
were prepared and many of them published alongwith the report of the Commission, which was issued
on October 30, 1979. In the course of its investigation,the ~ommission conducted 12 days of public hearings,
and Its staff compiled more than 150 separate depositions.
The report of the President's Commission was divided into three major sections: an overview, together
with the principal specific findings of the Commission
with respect to the causes of the accident; recommendations flowing from the findings and addressed to (1)the NRC, (2) the utility and nuclear industry, (3) the, training of nuclear plant personnel, (4) certain technical considerations, and (5) the health and safety ofplant workers and the general public; and a chronology of the accident with some further attribution of
causality. Highlights of each section are provided
below, together with the NRC's response to the Commission's recommendations and the President's statement about them.

Findings and Judgments
The Commission affirmed at the outset of its reportits basic conclusion that to prevent accidents as serious
as TMI in the future it will be necessary to effect "fundamental changes" in the organization, proceduresand practices, and, "above all, in the attitudes of the
Nuclear Regulatory Commission and, to the extent
that the institutions we investigated are typical, of thenuclear industry." The need for a change in attitude in
NRC and in the industry is emphasized throughout theCommission's report. The Commission also declared at
the start that its findings do not, "standing alone," require a conclusion that nuclear power plants are inherently too dangerous to continue in operation or thatnew ones should not be built, but neither would theCommission propose that the nation "move forward
aggressively" in expanding commercial nuclear poweruses.
In its discussion of causality, the Commission identified the root problems as being "people-related,"rather than related to deficiencies in plant design orequipment (though these too were present and involved in the accident). The weaknesses identified were
not only the "shortcomings of individual human beings," but problems of structure and communication"among key individuals and groups." The Commission
asserted outright that the equipment involved at TMI
was good enough that, "except for human failures, themajor accident ... would have been a minor incident:' There was, the Commission found, a preoccupation with regulations as such, rather than withthe safety they are supposed to promote, and that
regulations as voluminous and complex as those in current effect were actually a negative factor with respect
to safety. A particular distortion cited by the Commission was the concentration on large-scale or "worst
case" hazards to the neglect of less consequential but
more probable scenarios. Thus "the break of a hugepipe ... [is] studied extensively and diligently,"reflecting the attitude that if the worst accidents canbe controlled there is little to fear from lesser events.
The Commission pointed out that TMI was the resultof a combination of minor equipment failures which is"~ikely ~? occur much more often than the huge acCIdents, and that successful handling of minor
failures is usually going to depend more on quick andappropriate human reaction, in contrast to thenecessarily automatic and extremely fast response of
equipment to sudden, large-scale accidents. The Commission urged on the NRC and industry a newfoundrecognition that "human beings who manage andoperate the plants constitute an important safetysystem."On the subject of operator error at TMI the Commission noted that the training of TMI ope~ators (andthat of reactor operators in general) was "greatly deficient" in that it did not prepare them for dealing withthe extraordinary, with "something as confusing" as

36

the conditions created by multiple equipment failures.
Moreover, the TMI- 2 control room design was lacking
in many ways, "the control panel is huge, with hundreds of alarms, and ... some key indicators placed ..
. where the operators cannot see them." (More than
100 alarms were in fact activated in the early stages of
the accident, and, while the pressure and temperaturein the reactor coolant registered in the control room,
there was no indication to the operators that the combination of the two meant steam was forming.)
Altogether the design of the room and its gauges andequipment gave little attention to "the interaction between human beings and machines" and "ignored theneeds of operators during a slowly developing small
break accident." Some members of the Commission
favored a complete modernization of the control
rooms of a TMI design, and all of them agreed that "arelatively few and not very expensive improvements inthe control roorp could have significantly facilitatedmanagement of the accident." Thus the Commission
found that, while inappropriate operator action was amajor factor in the TMI accident, a number of deficiencies on the part of the utility, its suppliers, and the
NRC-in training, in procedures,. in control roomdesign-and the failure to recognize these deficiencies
and to learn from previous experience were amongmajor contributing causes. Despite its findings as tothe proximate and contributing causes of the TMI accident, and its judgment that the potential for suchlapses could and should have been anticipated byvarious principals involved, the Commission expressed
its conviction that, given all the deficiencies cited, "anaccident like Three Mile Island was eventually inevi ta ble. "Regarding the severity of the accident's impact on
public health, the President's Commission determinedthat actual releases of radiation at TMI "will have anegligible effect on the physical health of individuals,"
and that the major health effect of the accident wasmental stress. As to the possibility of an eventual TMIradiation-induced cancer occurring among the exposed population, it found that there will be "either nocase of cancer or the number of cases will be so smallthat it will never be possible to detect them." Themental stress experienced by people near the facilitywas "quite severe," however. The Commission ascribed this to several factors, especially the extensive
speculation by public officals during the first week of
the accident on how serious it could become andwhether evacuation of the population should or would
take place. Concerning the effect of news media
coverage during this time-its 'speculations, selectionsof items to cover, and general tone-the Commission
decided that there was "overall, a larger proportion ofreassuring than alarming statements in the coverage,"and the news media "did not present only 'alarming'
views, but rather views on both sides," although a"few newspapers . . . did present a more frightening

and misleading impression of the accident." Thesevere stress was short-lived, the Commission concluded, and was worst among people living within fivemiles of the plant and in families with young children .
The damage to the facility itself was very extensive
and, in the words of the report, the "ongoing cleanup
operation at TMI demonstrates that the plant was inadequately designed to cope with the cleanup of adamaged plant. The direct financial cost of the accident is enormous. Our best estimate puts it in a range
of $1 to $2 billion, even if TMI- 2 can be put back into
operation. (The largest portion of this is for replacement power estimated for the next few years.) And
since it may not be possible to put it back into operation, the cost could be much larger."
The Commission felt it an important part of theirtask to ascertain not only how bad the TMI accident
was but how bad it might have been. It posed thequestion to itself, "What if one more thing had gonewrong?" Among the possibilities considered waswhether a hydrogen or steam explosion could havebreached the reactor vessel and also the containment.
(That a nuclear explosion might have done so was notconsidered because, with the slightly enriched fuel used in a reactor, such an explosion is not a possibility.)
Several scenarios potentially leading to the rupture ofcontainment and release of massive amounts of radiation from the plant were studied. Of particular concern was the potential release of radioactive iodine
which might enter the food chain. (There was only a
trace off-site release of iodine from the actual TMI accident.) Some scenarios led to a better outcome than
the actuality, and two or three would have resulted in
more severe core damage than occurred and even a
melting of the core. However, the Commission
reported that-within the limits of current engineering knowledge of the interaction of molten reactor fuelwith concrete, steel and water-its calculations show"that even if a meltdown occurred, there is a highprobability that the containment building and thehard rock on which the TMI- 2 containment building
is built would have been able to prevent the escape of alarge amount of radioactivity." Being less than absolutely sure of this conclusion, the Commission urged
more research into this vital but murky area of severe
core damage and its worst plausible effects. The Commission averred that, whether or not TMI came close
to becoming catastrophic, "accidents as serious as TMI
should not be allowed to occur in the future,"
although "we must not assume that an accident of thisor greater seriousness cannot happen again, even if thechanges we recommend are made." The latter factargues strongly for the need to be prepared to deal
with the aftermath of such accidents.
The next focus of Commission scrutiny, closely
related to its last cited observation, was the matter of
emergency preparedness among the various govern-

37

mental elements involved at TMI. The Commission
judged that the plans made by these agencies were notadequate and that their responses to the emergency
were not satisfactory. It found problems associatedwith having multiple jurisdictions respond to a radiation emergency and an "almost total lack of detailedplans" in the local communities around TMI. The
report noted that when "prompt defensive steps arenecessary within a matter of hours, insufficient advance planning could prove extremely dangerous."
The Commission advocated centralization of emergency planning and response in a single Federal agency
which would maintain close coordination with Stateand local authorities and draw upon Federal and otherexpertise as the need arose. The report also criticized
the NRC siting policy with respect to nuclear facilities
and its requirement that reactors be located in a "lowpopulation zone," or LPZ, where protective action
could be taken in the event of an accident. The Commission found "this concept implemented in a strange,
unnatural and roundabout manner," with dimensions
predicated on only a very serious hypothetical accident accompanied by very large doses to the population. (The NRC discontinued use of the LPZ in its
siting requirements prior to publication of the Commission report.) The Commission proposed that avariety of possible accidents be considered in site
evaluation-particularly the smaller scale accident
with the higher probability of occurring-and protective action appropriate to each sector of the affectedpublic be built into emergency plans for a facility.
Also, State and local agencies must be ready to respond"once information is available on the nature of an accident and its likely levels of releases."
At TMI the emergency response "was dominated by
an atmosphere of almost total confusion," the reportstated, with "lack of communication at all levels."
On the subject of public and worker health and safety, the Commission noted that, in setting standards forworker exposure to radiation in licensed facilities, in
its plant siting and other health-related decisions, the
NRC "is not required to, and does not regularly seek"
advice or review from other Federal agencies, such as
HEW or EPA, concerned with health and radiation.
Emergency plans did assign responsibilities to theseagencies, as well as to DOE and NRC, in their
response to the TMI accident, but, the Commission indicated, the plans were so poor that ad hoc arrangements had to be made and coordination improvised.
In addition, the Commission found that the Stateagencies with responsibilities for public health did nothave adequate resources "for dealing with radiation
health programs related to the operation of TMI." Its
recommendations on these matters appear later in thischapter.
On the issue of whether the public's right to information during the accident was well served, the Commission concluded that it was not. It found "serious

problems with the sources of information, with how
this information was conveyed to the press, and alsowith the way the press reported what it heard." Earlyin the accident the utility tended to minimize thehazards, according to the report, while later on the
NRC "was the source of exaggerated stories." In particular, the Commission noted, "official sources would
make statements about radiation already released ...
that were not justified by the facts-at least not if thefacts had been correctly understood. And NRC was
slow in confirming good news about the hydrogen
bubble. On the other hand, the estimated extent of thedamage to the core was not fully revealed to the
public."A separate problem concerned the way facts were
presented to the press. It seemed that those who briefed the press either lacked the technical knowledge toexplain the events transpiring or, when they did
understand what was happening, they tended to speak
in a technical jargon the press could not understand.
Moreover, the report stated, "The press was further
disturbed by the fact that, in order to cut down on theamount of confusion, a number of potential sources of
information were instructed not to give out information. While this cut down on the amount of confusion,
it flew in the face of the long tradition of the press
checking facts with multiple sources." As mentioned,
the Commission concluded that, with a few notable
exceptions, the media "generally attempted to give abalanced presentation which would not contribute toan escalation of panic." (The Commissioner who wasresiding in Middletown, Pa., during the accident did
not concur in that judgment; see "Supplementary
Views," below.) A serious impediment in the conveying of accurate and complete information to the public
was that "even personnel representing the major national news media often did not have sufficient scientific or engineering background to understandthoroughly what they heard, and did not have avail-

NRC trailers at Three Mile Island used by the investigation teamof the Office of Inspection and Enforcement and by the TMI support staff of the Office of Nuclear Reactor Regulation. Staff fromthese and many other NRC offices were on duty from time totime-many on a voluntary basis-during the first few weeks after
the accident. Later, plans were made for office space in Middletown, Pa., for a long-term stay of some NRC staff.

38

able to them people to explain the information." Thisapplied particularly to the reporting of radiation
releases, when numbers told the public nothing of theseverity-or insignificance-of the releases. "Many of
the stories were so garbled as to make them useless as asource of information."
Turning to an assessment of the NRC, the Commission took note that "when NRC was split off from theold Atomic Energy Commission, the purpose ... was toseparate the regulators from those who were promoting the peaceful uses of atomic energy." But theCommission found "evidence that some of the old promotional philosophy" persists in the regulatory practices of the NRC, and "evidence ... that the NRC has
sometimes erred on the side of the industry's convenience rather than carrying out its primary mission ofassuring safety." In both the NRC's licensing and its
inspection and enforcement activities, the Commission
found "serious inadequacies."
The NRC licensing criteria and general approachwere found exceptionable in several key respects:• The application of a "single failure" criterion in
the licensing process and the failure to analyze theconsequences of a breakdown in two systems
occurring independently (as happened at TMI).
• The inappropriately sharp distinction between"safety-related" components and "nonsafetyrelated," and the exemption of the latter from thestringent requirements applied to the former.
(The report proposes instead •. a system ofpriorities as to how significant various components . . . are for the overall safety of theplant.")• The apparent assumption that plants can be made"people proof," and insufficient attention tooperator training and operating procedures.
• The licensing of plants when relevant safety issues
remain unresolved.
• Insufficient attention to the "ongoing process of
assuring nuclear safety," as exemplified by NRC's
categorization of a safety issue as a "genericproblem," thereby relieving the licensee of
responsibility for resolving the issue before licensing. (The report suggests there is evidence that"the labeling of a problem as generic may provide
a convenient way 6f postponing decision on a difficult question.")• A reluctance to apply new safety standards topreviously licensed plants. (The report cites thisas an instance of the "old AEC attitude" influencing NRC judgments and finds no evidence that"the need for improvement of older plants was
systematically considered prior to Three Mile
Island.")• The tendency of a detailed and voluminous bodyof regulations "to focus industry attention narrowly on the meeting of regulations rather thanon a systematic concern for safety." (The Commission felt that, in some instances, certain
regulations may-because of the way rate bases
are decided-have served to deter utilities andsuppliers from initiating safety improvements.)• The voluminous NRC inspection and enforcement
manual, so extensive that "many inspectors do notunderstand precisely what they are supposed todo." The Commission also found that sometimes
inspectors have had difficulty getting theirsuperiors "to concentrate on serious safety issues,"
and also that incidents reported by licensees"tended to concentrate on equipment malfunction" while "serious operator errors have not been
focused on."• The lack of a systematic method for evaluating industry experience and to look for patterns that
could warn of the presence of a basic problem,
and a failure to use monetary fines to full effect.
• A heavy preoccupation in NRC with the safe
operation of equipment to the neglect of "peopleoriented" concerns, resulting in lack of attention
to the operating procedures and .. an almost totallack of attention" to the interaction between
human beings and machines.
With respect to the NRC's response to the TMI accident, the Commission stated that it was "extremely
critical of the role the organization played," citing the"serious lack of communication among the commissioners, those who were attempting to make the decisions about the accident in Bethesda, the field offices,and those actually onsite." The Commission questioned the suitability of the collegial structure of NRC,
with five equipollent Commissioners, to manage an
emergency and it found the "precise role" of the Commissioners unclear. In addition, the President's Commission observed that the "huge bureaucracy [NRC]."is highly compartmentalized with insufficient communication among the major offices," and it saw no"effective managerial guidance from the top," butrather "some of the old AEC promotional philosophy
in key officers below the top." The Commission also
cited the unnecessarily strict procedural rules within
NRC which inhibited free communication among the
NRC Commissioners and between them and the staff,
In conclusion, the President's Commission determined that, despite in-depth studies and critiquesfrom within and outside the agency, there is "no wellthought out, integrated system for the assurance of
nuclear safety within the current NRC," For all of thereasons discussed, the Commission recommended a"total restructuring of the NRC," making the agency
part of the executive branch, headed by a single administrator chosen from outside the NRC, with thefreedom to "reorganize and bring new blood into the. , . staff. This new blood could result in the change of
attitude that is vital for the solution of the problems of

==~~~~================================~==~====39

the nuclear industry." This and other Commission
recommendations are treated below (see "Recommendations and Responses"), together with the NRC's
response to each,· as well as the PreSident's statementon the Commission report.
With regard to the utility, the President's Commission felt that the necessary "management qualifications and attitudes" for conducting nuclear power
operations were not given sufficient attention by theparent corporation whose subsidiary ran TMI. The
Commission found "a divided system of decisionmaking within [the parent company, General Public
Utilities Corporation] and its subsidiaries. While theutility has legal responsibility for a wide range of fundamental decisions, from plant design to operator
training, some utilities have to rely heavily on the expertise of their suppliers and on the Nuclear
Regulatory Commission. Our report contains a
number of examples where this divided
responsibility, ... [as] in the case of TMI, may have
led to less than optimal design and operating
practices." The report notes that the design of the TMI
control room "seems to have been a compromiseamong the utility, its parent company, the architectengineer, and the nuclear steam system supplier (withvery little attention from the NRC)." Operator training afforded the best example of the effects of divided
responsibility, however. The utility has the legal
responsibility for training operators and supervisors,
but the TMI licensee did not have the expertise to conduct training by itself, so it contracted with the supplier of the nuclear steam system to do some portionsof the training. The latter company had no responsibility for the quality of the total program, and coordination between it and the licensee was "extremely
loose." The simulator employed in the program givenby the reactor supplier differed "in certain significant
ways" from the actual control console at TMI and, inany case, it was not programmed to reproduce theconditions faced by the TMI operators on March 28.
The Commission believed that "the role that the NRC
plays in monitoring operator training contributes little
and may actually aggravate the problem." The NRC's"fairly routine licensing examinations" and limited
spot-checking of requalifications exams (administeredby the utility) "may be perpetuating a level ofmediocrity," since the utility tends to equate the passing of the NRC exam with satisfactory operator training. The report was again very critical of operatingprocedures at TMI and the corresponding deficiencies
they produce in the operators' training. Commission
analysis of TMI management "raises the serious question of whether all electric utilities automatically havethe necessary technical expertise and managerial
capabilities for administering such a dangerous hightechnology plant." ConcllJding that they do not, theCommission recommended higher standards oforganization and management that a company mustmeet before receiving an operating license.
Recognizing that "recommendations as sweeping as
ours will take a significant amount of time to implement," the Commission unanimously voted that "theNRC or its successor should, on a case-by-case basis,
before issuing a new construction permit or operating
license: (a) assess the need to introduce new safety improvements recommended in this report, and in NRC
and industry studies; (b) review, considering therecommendations set forth in this report, the competency of the prospective operating licensee tomanage the plant and the adequacy of its training program for operating personnel; and (c) condition licensing upon review and approval of the State and local
emergency plans."
Expressing its "overwhelming concern about someof the reports" from other TMI investigations, andwarning that proposed improvements carried out in a"business as usual" atmosphere will not suffice, thePresident's Commission concluded the Overview,
stating:"We believe that we have conSCientiously carriedout the mandate of the President of the United States,
within our limits as human beings and within thelimitations of the time allowed us. We have not founda magic formula that would guarantee that there will
be no serious future nuclear accidents. Nor have wecome up with a detailed blueprint for nuclear safety.
And our recommendations will require great efforts byothers to translate them into effective plans."
The Commission reaffirmed the need for fundamental change, charging that "unless portions of the industry and its regulatory agency undergo fundamental
changes, they will over time destroy public confidence
and hence, they [emphasis theirs] will be responsible
for the elimination of nuclear power as a viable sourceof energy."

Supplemental Views
A number of Commissioners published comments of
their own as supplements to the overall report of the
President's Commission.
The Chairman and five other Commissioners
cosigned a statement taking note of the fact that they
had supported a recommendation, which failed ofadoption by the full Commission, that "no new work
authorization permits or constructions permits should
be issued until such time as the NRC or its successor
had adopted siting guidelines" consistent with therecommendation, which was adopted (unanimously),
calling upon NRC to review its siting criteria (see"Recommendations and Responses;' recommendation
number 6, below).
Governor Bruce Babbitt of Arizona took up the matter of utility capability to operate nuclear power
plants and gave his view that, while the "Commission

40

Here a workman prepares to enter a contaminated area by donning a suit of protective clothing. He is careful to tape his ankles toseal the area where the pant leg of the suit joins the overshoes.

Workmen about to leave the plant are carefully monitored with apancake-type Geiger counter to insure that they have not been
contaminated.

has clearly addressed the institutional shortcomings of
the Nuclear Regulatory Commission, it has not addressed the institutional problems of the industry."
The Governor expressed his belief that "this is one area
where fewer entities with more depth and expertisemight be justified for sake of public health and safety."
The Commissioner also mentioned the possibility thatcertain facts known to TMI management on the firstday of the accident had not been conveyed in timely
fashion to the NRC and State officials, an issue which
merits further investigation.
Commissioner Russell W. Peterson, President of the
National Audubon Society, reaffirmed his endorsement of the recommendation noted above as having
the support of six Commissioners, namely, that no newlimited work authorizations or construction permits
should be issued until the NRC siting requirements
were changed. The Commissioner also felt that the
President and Congress should "involve highly
technically qualified critics of nuclear energy safety"
in the continuing appraisal of nuclear safety called forby the Commission. He also urged serious study of
nuclear waste disposal. Commissioner Peterson finallystated his conviction that a "much more serious accident" than TMI was going to occur somewhere atsome time, because of the complexity of the technology
and human limitations, and therefore he called for thedevelopment by the government of a "strategy which
does not require nuclear fission energy."

Commissioner Thomas H. Pigford, Chairman of the
Department of Nuclear Engineering at the Universityof California (Berkeley), issued a lengthy supplement
to the report, setting forth, among others, the following observations:• The report's stress on the need for more emphasison people and less on equipment has obscured the"very important fact" that, despite the crucial
errors of people, the safety equipment did indeedfunction to achieve its purpose; and despite thefailures of equipment-the stuck valve and theleaks in the gas vent system-the overall system
was good enough that, absent the effects ofhuman error, the accident would have been aminor incident. Staff analyses show that even if
all the fuel cladding had oxidized and even if fuelmelting or meltdown had occurred, the containment would have stood up and the public would
have been protected.
• Systems of equipment at TMI performed better
than expected; earlier assumptions would have
led to far greater core damage and radiation
release to the containment than what actually occurred.
• TMI has revealed to all a number of remedies andimprovements to be made, but there "seems to besome unwillingness to recognize that many ofthese remedies are already being implemented.

=====================================================41
The NRC and the nuclear industry have taken
and are taking steps ... The problem with 'attitudes' emphasized in the Commission's reportmust refer largely to pre-TMI attitudes."• More emphasis is needed on analysis of and planning for small break accidents, but "the possibility of an accident of this type was known and had
been analyzed and predicted prior to the TMI-2
accident." Thus the facts of the present investigation provide no basis for concluding that reactors
are unsafe.
• Since the attitudes of various persons and groups
were not directly examined prior to the TMI-2
accident, valid conclusions can only be drawn on"actions taken, i.e., problems addressed and not
addressed, regulations issued and complied with,and the occurrence of events that reflect upon theadequacy of these processes." It is "more constructive to assume that attitudes are symptomatic of ... forces at work in the system, and itis those forces which must be addressed." It is theapparent failure of the system to assimilate lessons
from plant experience and to incorporate up-todate technology-in control room design, for example-that constitute "a more manageable andappropriate focus for the overall conclusion ofthis Commission. I believe that such technology
is . . . or will be used by the industry and thatchanges ... will be effected, not merely tosatisfy critics or to demonstrate attitudinal
penitence, but on the basis of sound judgment
resting on sound data."• The NRC must deal with the question of how
much cost and delay is justifiable to realize agiven increment in safety, and efforts to balancecosts and benefits should not be considered evidence per se of a promotional philosophy. Bothoverreaction and inaction in this area carry social
costs which must be weighed.
• While it is "confusingly" referred to as a "single
failure" criterion, the NRC licensing process applies a criterion which assumes at least three
failures: any credible component failure (1) inwhich all internal or all external power supply is
lost, with (2) the additional failure of a single active component which (3) is the component whose
failure causes the most serious aggravation of theaccident.
• In the analysis of postulated accidents, there is no
assumption that an active "nons-afety-related"
item will not fail; it was not a preoccupation witha safety-related item list that proved inadequate
in the analysis of TMI, but a failure to take into
account lack of operator training and deficient
operating procedures and practices.
• The finding that there is no systematic backfitting
review of older plants "appears to take too little

account" of NRC's Systematic Evaluation Program (SEP), initiated more than three years ago;
progress in some areas, such as upgrading
emergency plans, does appear to have beensomewhat slow.
• The Commission's appraisal of NRC inspection
and auditing of licensee compliance "calls for
NRC to do more of what it already does and to do
it better." Resident inspectors have been at someplants for more than a year, and unannounced
on-site inspections "appear to be so frequent as tobe commonplace." It is "clearly impractical" forthe NRC to undertake substantial independent
testing of construction work and cease to rely on
testing done by the utility.
• A lack of quantified safety goals is a major problem in the NRC regulatory rationale, and its
failure to set priorities leads to a disproportionate
commitment of resources and efforts to sometimes
marginal concerns. A large portion of the NRC
management and staff are lacking practicalexperience in designing and operating the equipment they regulate, and too many requirements
are unsupported by valid technical backup andvalue-impact analysis (an "overwhelming emphasis on conservative models and assumptions").
There is an insufficient exchange of information
between NRC and industry because of the "adversary approach" existing between them, and NRC
does not carry out the kind of systematic analysis
of operating data that would disclose significant
trends and patterns.
• There was not sufficient time allowed for acareful review of the President's Commission staff
reports on which Commission findings werebased (some were still incomplete when the final
report was issued), and there were "several partsof some key staff reports with which I cannot
agree, particulary the staff report on the NRC."
There was unqualified acceptance in that reportof testimony which was unconfirmed and uncorroborated, "an indicator of insufficient balance"in the staff investigation of the NRC. The staff
report also "relies to a considerable extent upon
excerpts from a book," without establishing theauthor's qualifications or taking his testimony.
The Commissioner stated, "In my view,the . . .. book does not express a comprehensive,
accurate and balanced knowledge of the NRC
and of the nuclear industry."• Criticism of the NRC "should not obscure the central issue that primary responsibility for nuclear
safety lies with the utility, shared to a large extent
with the equipment suppliers and the architectengineers. This also reflects my view of the
responsibilities for the TMI-2 accident."

42

Commissioner Anne D. Trunk, a resident of Middletown, Pa., located about three miles from the TMI
station, took exception to the Commission finding
regarding the news media's treatment of the accident
and its effect on the mental state of the people living
near the facility. (Mental stress was identified by theCommission as the "major health effect" of the accident.) Commissioner Trunk, affirming that she spoke
for herself "and a majority of her circle of citizens who
lived through the TMI accident," stated:"The report concluded that the errors and sensationalism reported by the news media merely reflected
the confusion and ignorance of the facts by the official
sources of information. It further concluded that thepress did a creditable ('more reassuring thanalarming') job of news coverage.
"In fact, these conclusions are not generally supported by the staff reports. There were reliable news
sources available. Too much emphasis was placed on
the 'what if rather than the 'what is.' As a result, thepublic was pulled into a state of terror, of
psychological stress."

Stacks of lead ingots were sent to the TMI acci~ent site from industry groups and national laboratories respondmg to a general request from the NRC. The lead was used in various/arts of theplant for radiation shielding during observation an measurement

The Commissioner called for a self-evaluation on
the part of the news media. She also noted that she
could not support a moratorium on the issuance of
new construction permits because "it was not shownhow this could result in a safer plant at TMI nor attainhigher standards of safety and performance by the industry." Instead, the Commissioner recommended a
defined period within which the parties concerned
would be charged to act upon the Commission's
recommendations, and a separate probationary
operating period for the licensee at TMI.

j;>

Recommendations and Responses
Starting below and in the pages that follow, thespecific recommendations of the President's Commission-concerning the NRC, the utility and its suppliers, the training of operating personnel, a technicalassessment, and both worker and public health andsafety-are set forth in the left-hand column, with the
response of the NRC to each recommendation set forthin the right-hand column.

taking. In the weeks following the accident, however. it was determined that site radiation levels did not require all of the leadand much of it was returned to the donors.

=================================================43tance, the Chairman noted, was the need for "prompt
and positive assurance that the technical and management competence of all licensees is sufficient to operate
nuclear power plants safely and to respond effectively
to emergencies." Expeditious action would be taken inthis area. Reporting that four of the five NRC Commissioners felt that effective reform could and shouldbe accomplished within the existing agency, theChairman also conveyed disagreement "with theoverall thrust of the President's Commission recommendations to lessen the role of NRC in responding toemergencies and providing emergency information to
the public." Estimating that it would take several
months to develop the new or improved safety objectives and detailed criteria for implementing them, theChairman disclosed that "we have decided that newplants will not be licensed until we have developed therequired criteria."

In forwarding the NRC comments to Dr. Frank
Press, Director of the Office of Science andTechnology Policy, Executive Office of the President,
Chairman Joseph Hendrie expressed a number ofgeneral comments on behalf of the Nuclear Regulatory
Commission. (Two NRC Commissioners added supplementary remarks, cited at the close of this section.)
The Chairman stated that, from NRC's own reviews of
the accident, "we have generally found that the actions recommended by the President's Commission in
the areas of human factors, operational safety,
emergency planning, nuclear power plant design andsiting, health effects, and public information arenecessary and feasible. He affirmed that changes
taken and intended by the NRC are in conformity withthe recommendations of the President's Commission,and that some changes under consideration would go
beyond those recommendations. Of particular imporU

PRESIDENT'S COMMISSION (PC) RECOMMENDATIONS ON THE NRC

PC RECOMMENDATIONSNRC RESPONSES(1) NRC SHOULD BE RESTRUCTURED AS A
NEW INDEPENDENT AGENCY IN THE EXECUTIVE BRANCH. The present five-member Commission should be abolished, and a single administrator
appointed by the President, with advice and consent of
the Senate, to serve at the pleasure of the
President. The administrator should be fromoutside NRC and should be given substantial
discretionary authority in managing the agency.

(1) Four of the five Commissioners felt that the objectives of the President's Commission could be accomplished by reforms effected within the existing
structure. It is desirable to have the statutory
authority to delegate management responsibilities
to a single Commissioner in event of an emergency. Clarifications in the law could remove
ambiguity of the Chairman's authority, as wellas that of the Executive Director for Operations. NRC
has adopted a new "policy planning program guide"mechanism and is studying new modes of Commission
involvement in developing key safety policy.

...

...

...

(2) AN OVERSIGHT COMMITTEE ON NUCLEAR
REACTOR SAFETY SHOULD BE ESTABLISHED.
Its purpose would be to examine, on a continuingbasis, the performance of the agency and the industry
in resolving important public safety issues related
to nuclear power plants and in exploring theoverall risks of nuclear power. Membershipup to 15 in number-would be drawn from thefields of public health, environmental protection, emergency planning, energy technology andpolicy, nuclear power generation, and nuclear safety;
one or more State governors and members of thegeneral public would serve on the committee, which
would report to the President and Congress annually.
...

" "

...

" "

(2) Although this call for an oversight committee is tied to the recommendation for a new
executive branch agency, this proposal should
be examined on its own merits. Such an oversight or public advisory committee might contribute to the interaction among the Federal
Government, States, utilities, public interest groups.
and the general public on the controversial
issues related to nuclear power.

...

"

...

44

PC RECOMMENDATIONSNRC RESPONSES(3) THE ADVISORY COMMITTEE ON REACTORSAFEGUARDS (ACRS) SHOULD BE RETAINED,IN A STRENGTHENED ROLE, TO CONTINUEPROVIDING AN INDEPENDENT TECHNICALCHECK ON SAFETY MATTERS. The staff of the
ACRS should be augmented, and its public healthexpertise especially improved. The ACRS would
choose which licensee applications to review, and it
would have a statutory right to intervene in hearings
as a party. It should be authorized to raise any safetyissue in a proceeding, give reasons and arguments for
its views, and require formal response by the agency
to its submissions. Any ACRS member would be
exempt from subpoena in any proceeding in which hehas not previously appeared voluntarily or madean individual written submission. ACRS should have
similar rights in rulemaking proceedings and power
to initiate such a proceeding to resolve any
generic safety issue it wishes.

(3) NRC endorses a strengthened role for the ACRS
and the recently initiated ACRS Fellows Program
should reinforce its analytic resources. But thestrength and value of the independent ACRS reviews
derives from the collegial interaction of its members;
adding staff beyond reasonable needs will not contribute much to that strength. NRC has supported
legislation which would enable the ACRS to choose
applications for review. The proposed right to intervene may not be appropriate for a part-time adVisory
body; it would require a new ACRS legal staff
and active involvement in hearings could severelycompromise the independence and collegial natureof the committee. The ACRS can now recommend
rulemaking to the NRC, but whether it should be
able to mandate a proceeding needs and will begiven further consideration. In general, NRC agrees
that ACRS views warrant prompt response by the
NRC staff. Comments on the matter have been requested from the ACRS.

*

*

*

(4) INCLUDED IN THE AGENCY'S GENERAL SUBSTANTIVE CHARGE SHOULD BE THE REQUIREMENT TO ESTABLISH AND EXPLAIN SAFETYCOST TRADE-OFFS. Where additional safety improvements are not clearly outweighed by cost considerations there should be a presumption in favor of thesafety change. The agency should be relieved of "anyunnecessary responsibilities that are not germane tosafety." In particular, operator and supervisor
licensing should be upgraded, and accreditation oftraining centers required; a definition of "safety matters" should be formulated which is broader than thepresent inventory of "safety-related items"; an emphasis on examination of overall plant design and performance, from a systems engineer's standpoint, isneeded, with attention to multiple failure potential,
control room design, instrumentation; research with abroad scope that includes public health and which exploits all scientific knowledge available should becoordinated with the regulatory process.

*

*

*

*

*

(4) NRC has not, in the past, clearly articulated its
policy on the effect of costs on safety decisions.
Some safety-cost tradeoffs are presently authorized,
e.g., value-impact analyses performed for proposed
regulatory requirements or in research planning. A
better articulation of NRC policy is needed. It is believed that benefits and detriments can be sufficiently
quantified to aid in decision-making, and it isagreed that, in general, some sort of safety-cost tradeoffs are at least implicit in a regulatory system thatconcedes that a goal of zero risk is impossible of attainment. The reality should be made explicit. But NRC isin complete accord that in all comparative judgmentsof this kind there should be a presumption in favor ofsafety. NRC will seek views of the Congress, otheragencies arid the public in developing an explicit
policy statement. Legislation may eventually bedesirable for the definitive policy expression. Legislation would be required to divest NRC of its non-safety
responsibilities, and the prospect raises problems in
the area of nuclear exports. The Commissioners are
not in agreement now on the best course of action. As
to operator and supervisor training, a study is under
way as to the options for NRC involvement andoperator licensing requirements are being upgraded.
The broadening of the definition of safety-related matters is apriority, including both equipment andhuman factors, and the interaction of safety- and nonsafety grade equipment is under study. Control roomdesign, overall plant design, and safety research are allundergoing reevaluation, and flexibility in assuring
maximum application of scientific knowledge will be
pursued.

***==~===============================================45

PC RECOMMENDATIONSNRC RESPONSES(5) RESPONSIBILITY AND ACCOUNT ABILITYFOR SAFE POWER PLANT OPERATIONS, INCLUDING THE MANAGEMENT OF A PLANT
DURING AN ACCIDENT SHOULD BE PLACEDON THE LICENSEE IN ALL CIRCUMSTANCES.
Thus the competence of licensees to meet this obligation must be assured, and the agency should impose higher standards or organizational and managerial capability, especially confirming the "integration of decisionmaking" in the company licensed
to construct or operate a plant; the necessary range ofexpertise; financial capability; quality assurance;
operator and supervisor performance; surveillance
and maintenance practices; and thorough analysis
and reporting of unusual events.

(5) NRC fully agrees and has begun actions to
upgrade standards and requirements to assure technical competence of licensees. The objective willbe "to minimize accident occurrence and maximize
proper response to accidents." Licensee performance
will be subject to more frequent periodic reviews, involving licensee's top management. More immediate
and decisive action is being contemplated (see response to recommendation 2 under "Commission
Recommendations on the Utility," below).

*

*

*

(6) THE AGENCY SHOULD BE REQUIRED, TO
THE MAXIMUM EXTENT FEASIBLE, TO LOCA TE NEW POWER PLANTS IN AREAS REMOTE
FROM CONCENTRATIONS OF POPULATION.
Siting determinations should be based on technical assessments of various classes of accidents that
can take place, including those involving releases oflow doses of radiation.

*

*

*

(7) THE AGENCY SHOULD BE DIRECTED TO
INCLUDE IN ITS LICENSING REQUIREMENTS
PLANS FOR THE MITIGATION OF THE
CONSEQUENCES OF ACCIDENTS, including thecleanup and recovery of the contaminated plant. The
agency should be directed to review existing licenses
and to set deadlines for accomplishing any necessary
modifications.
***(8) BEFORE ISSUING A NEW CONSTRUCTIONPERMIT OR OPERATING LICENSE, THE NRCSHOULD DO THE FOLLOWING ON A CASE-BYCASE BASIS: assess the need to introduce the safetymeasures recommended by the President's Commission and in NRC and industry studies; review the competence of the prospective licensee to manage the plant
and the adequacy of operating personnel training; andmake licensing contingent upon review and approvalof State and local emergency plans.

*

*

*

*

*

(6) The NRC Siting Policy Task Force report undercurrent review by the Commissioners recommends
similar changes and goes beyond those proposed.
Radiation releases from small accidents will beconsidered in appraising these recommendations. For
the past five years, the Standard Review Plan has
excluded sites with high population densities, but
operating plants built before then may call for added
design features, power reduction, or shutdown.
***(7) The NRC Lessons Learned Task Force recommends similar action but goes beyond that proposed.
The staff has already implemented new requirements for system leakage and shielding and has
recommended operator training in core-melt accident
mitigation, as well as NRC rulemaking on required
design features to provide such mitigation.
***(8) NRC has decided that new plants will not belicensed until the required criteria have been developed. The NRC will: (a) review and correlate recommendations of the President's Commission, the ACRS,
the Congress, its own inquiries and others; (b) draw up
safety objectives corresponding with those recommendations; (c) develop plans by which to realize those objectives by action affecting NRC structure and procedure or by requirements placed on licensees; (d) impose such requirements on operating plants; and (e)impose such requirements on plants under construction. Deadlines will be associated with the last two
steps. Operator training will, as noted, be upgraded,
and a rule requiring approval of State and local
emergency plans prior to plant operation is being considered.
*

*

*

46

PC RECOMMENDATIONSNRC RESPONSES(9) THE AGENCY'S AUTHORIZATION TO MAKE
GENERAL RULES AFFECTING SAFETY SHOULD
INCLUDE THE FOLLOWING REQUIREMENTS:
that a public agenda be developed according to whichrules will be formulated; that the agency set deadlines
for resolving generic safety issues; that existing rules
be reevaluated periodically and systematically; thatrulemaking procedures be adopted which giveinterested persons a meaningful opportunity toparticipate, which ensure careful considerationand explanation of proposed rules, and which
provide for the application of new rules toexisting plants. In particular, proposed rules
should be accompanied by analyses of the issues involved and identification of relevant technicalmaterial. Interested parties and organizations
should have sufficient opportunity to assess andrefute technical evidence and findings, and final
rules should be fully explained, with responses
for principal comments received. If needed,
interim safeguards for operating plants affectedby generic safety rulemaking should be imposed,
and the possible need for retroactive applicationof new safety requirements to operating plants
should be examined.

(9) NRC publishes an agenda of rulemaking petitions,a report of regulations under development, advancenotices of proposed rulemaking in major actions, andproposed rules for comments. Analyses and discussionsof these are made public, and public meetings or hearing are held in cases of special importance. The means
for the public to petition NRC to issue, revise or withdraw a rule are provided, and proposed and final rules
sent to NRC Commissioners for consideration are accompanied by a staff paper dealing with the relevant
concerns, alternatives, benefits and detriments, andcomments received and their resolution. The process isbeing reevaluated for clarity, sufficiency of public attention, effectiveness in resolving safety issues. In
practice all new rules call for a judgment on backfitting to existing plants, but NRC is now considering
including the practice in the regulations. Deadlines forthe resolution of unresolved safety issues were set morethan a year ago, and, these issues are, by definition, themost significant of the generic issues. Other such issues
will be addressed by priority based on safetysignificance. The review of NRC regulations usually
has followed some specific event, such as a research
result, a petition for rulemaking or new technology,
with some exceptions in the area of transportationand safeguards. This will now change, with plans for
an initial review of regulations by June 1980, completion of relevant rule changes by 1982, and completionof a systematic review of all safety regulations by 1984.
The review cycle will be repeated thereafter every
five-to-seven years.

••(10) LICENSING PROCEDURES SHOULD
FOSTER EARLY RESOLUTION OF SAFETYISSUES BEFORE MAJOR FINANCIAL COMMITMENTS IN CONSTRUCTION CAN OCCUR.
The Commission recommends: (a) the reductionof duplicative consideration of issues in the several
stages of a plant's licensing history by assignment ofparticular issues (such as need for power) to some
single stage of the proceedings; (b) resolution of issues
that recur in many licensing actions by rulemaking; (c)combining construction permit and operating license
hearings whenever plans can be made sufficiently
complete at the construction permit stage; (d) an initial adjudication of a license application and appeal
to a board whose decision would be final, with no provision for subsequent appeal within the agency. Bothadjudicators and appeal boards would have a clear
mandate to pursue any safety issue it wished to; (e) the
creation of an "Office of Hearing Counsel" in theagency to participate in formal hearings as ~'an objective party, seeking to assure that vital safety issues are
addressed and resolved;' and empowered to appeal•••(10) The objective underlying this recommendation is shared by NRC, but it cannot make specificcomment on it at present. A report is pending
from a special advisory committee on its study ofan NRC rule which permits plant construction during
adjudication. The report may also have a bearing
on the NRC practice of permitting discrete, specificissues to remain open up to the operating license
stage and even beyond. (It can happen that a safetyissue cannot be settled without additional information, but that such information can be obtained byresearch, even as construction proceeds.) On November 2, 1979, the NRC suspended its rule by which reactor licenses become immediately effective following afavorable initial decision by a licensing board. No
license will become effective until the Commission
itself has had the opportunity to determine the relevance of TMI-related issues to the case. The assignment of single issues to specific stages of the process,and possibly combining construction permit andoperating license hearings, are matters in which NRC's
authority is unclear (the latter step would require new

47

PC RECOMMENDATIONSNRC RESPONSES"any adverse licensing board determination to theappeal board( and (f) a deadline on the resolution of
any specific safety issue left open in a licensing
proceeding.

statutory power). Even though it may be possible tocombine the two kinds of hearings, there must still be avehicle for verifying the design details, and that mustnecessarily be done late in construction whenengineering of the design is complete. Also, new information affecting the early construction permit decision can arise at any time. It is current NRC practice
to segregate recurrent issues for generic resolution
whenever possible. The recommendation that appeal
board decisions be made final NRC dispositions of applications for licenses would have the effect of removing the Commissioners (or Administrator) entirely
from a major dimension of nuclear regulation. As to
the mandate to pursue safety issues, the boards alreadyhave independent authority to pursue "serious
matters" and the exercise of the right is no longer
qualified by "sparingly" or "in extraordinary circumstances." The proposal that a new Office of Hearing
Counsel be created has a purpose which is not entirely
clear, but it might serve as an alternative to otherdevices for broadening public participation, such as
intervenor funding, and merits consideration. Plantspecific safety issues left open at the time of licensing
are now carried forward with clear deadlines as conditions on the operating license; NRC will consider
whether it should also be conditioned with deadlines
for resolution of relevant unresolved safety issues.

*(11) THE AGENCY'S INSPECTION AND ENFORCEMENT FUNCTIONS MUST RECEIVE INCREASED EMPHASIS AND IMPROVED MANAGEMENT. The Commission recommends: (a) an improved program for the systematic safety evaluation ofplants to assess compliance with requirements, todetermine whether new requirements should apply,
and to identify new safety issues; (b) systematic assessment of reactor operating experience to reveal any pattern of abnormal activilty and provide a measure ofoverall rises or declines in safety and a base for specificimprovements; (c) substantial penalties be levied on
licensees who fail to report new safety-related information or violate rules proscribing unsafe practices;(d) improved inspection and auditing of licensee compliance with regulations and unannounced onsite inspection; (e) periodic intensive and open review of
each licensee's performance in meeting license requirements and regulations; and (f) agency adoptionof criteria for revocation of licenses, for sanctions shortof revocation (e. g., probation), and for requiring immediate plant shutdown or other operationalsafeguards.

*(11) In 1977, NRC set up a Systematic Evaluation
Program (SEP) whose first phase called for review ofconditions at 11 older plants. Extension of this program to all operating plants is being considered. The
Interim Reliability Evaluation Program (see recommendation 4 under "Technical Assessment," below)
is also under consideration. In July 1979, NRC
created the Office for Analysis and Evaluation of
Operational Data to give broad coordination to major
program offices' assessment of operating experience;
licensees have also been required to establish operating
experience evaluation groups and to assess experienceof other facilities than their own. The industry has
created similar groups. The inspection and enforcement staff is being augmented with plant systems
analysts to conduct independent technical evaluations
and followup of licensee events, transients, and inspection findings. Potential generic problems andoperating experiences will be conveyed promptly tolicensees through Bulletins, Circulars, and Information Notices. Legislation to increase civil penalties imposed by NRC is pending before Congress, and thepossible use of probation status is under review within
NRC. The resident inspector program begun in 1977
has been expanded; at least two resident inspectors
will be assigned to each site in fiscal year 1981.
Licensee performance evaluations combined with

48

PC RECOMMENDATIONSNRC RESPONSES
assessments of licensee management control systems by
the Performance Appraisal Team will identify
marginal utility operations and provide prompt correction. Unannounced inspections are carried out by
NRC, but the need for these in light of the expanded
resident inspector program is problematic.

PRESIDENT'S COMMISSION (PC) RECOMMENDATIONSON THE UTILITY AND ITS SUPPLIERS
PC RECOMMENDATIONSNRC RESPONSES(1) THE NUCLEAR INDUSTRY MUST DRAMATICALLY CHANGE ITS ATTITUDES TOWARD
SAFETY AND REGULATIONS; IT MUST SET AND
POLICE ITS OWN STANDARDS OF EXCELLENCE, to ensure the effective management andsafe operation of nuclear power plants. It should
develop standards for management, quality assurance,and operating procedures and practices, and it should
conduct independent evaluations (perhaps through theInstitute of Nuclear Power Operations). It should
gather and analyze all power plant operating experience systematically, communicate information speedily to affected parties, and make needed changes on
realistic deadlines.

(1) NRC agrees that imRrovements and maintenanceof operational safety is a fundamental responsibility oflicensees. The NRC role should be to provide acceptance criteria, detailed guidance where necessary, andany incentives, needed to attain and sustain operational safety. NRC agrees with the other parts ofrecommendation 1 as well and feels the Institute of
Nuclear Power Operations may well be the right vehicle for independent evaluation, especially with regard
to important human factors. A statement ofunderstanding between the Institute and the NRC
should be executed within six months. In addition tocreating the Office of Operational Data Analysis and
Evaluation, the NRC has required each licensee toestablish an engineering staff capability to assess andfeed back pertinent operating experience. The intent isthat programs of NRC, industry, and vendors will becomplemented by and integrated with each licensee's
program to assure that intelligible analyses ofoperating experience reach all reactor operators andplant technical support staff. A proposed rulemaking
by NRC would require plant shutdown by a licensee
upon discovery of human or operational errors thatcause important safety systems to be inoperative.

* * *(2) ALTHOUGH RESPONSIBILITY FOR SAFETY
LIES WITH THE TOTAL ORGANIZATION OF
THE PLANT, EACH NUCLEAR POWER PLANT
COMPANY SHOULD HAVE A SEPARATE SAFETY
GROUP THAT REPORTS TO HIGH-LEVELMANAGEMENT. The group's assignment would be toevaluate procedures and general operations regularly
from a safety perspective, to assess quality assurance
programs, and to develop continuing safety programs.

*

*

*

(2) Although NRC has taken action to augment on-site
technical support capability with shift technical advisors and operations evaluators at each plant, it isconsidering a requirement that would expand the staff
for on-site safety surveillance by all licensees. A groupof technical specialists would be assembled with nodirect operating responsibilities to distract them fromday-to-day attention to safety; it would report tosenior management independently of the power production staff. NRC is also considering a requirement
for licensees to improve their systems for independent
verification of operational safety by means ofautomatic system status monitoring and personalverification as well.

====================================================49
PC RECOMMENDATIONSNRC RESPONSES(3) INTEGRATION OF MANAGEMENT RESPONSIBILITY AT ALL LEVELS MUST BE ACHIEVEDCONSISTENTLY THROUGHOUT THIS INDUSTRY. There must be a single accountable organization with the requisite expertise to take responsibility for the integrated management of the design,
construction, operation, and emergency response
functions of nuclear power plant operation. Without
such demonstrated competence, a company should not
qualify for an operating license. At the design stage,
the utility can either contract for a "turn-key" plant, afully operational plant delivered by the vendor orarchitect-engineer, or the company can assemble expertise capable of integrating the design process. In
either case, it is critical that knowledge gained during
design and construction of the plant be transferred effectively to those responsible for operating the plant.
Clear procedures, responsibilities, and communication serve to ensure accountability and are especially
important in the event of a crisis.

(3) NRC has recently surveyed and is (studying thetechnical resources available to each power reactor
licensee. It is developing new criteria by which tojudge the competence of licensees to operate nuclear
facilities and expects to promulgate them by April
1980. NRC agrees emphatically that there is a need forclear definition of roles and responsibilities and has required that licensees for operating plants provide thesekinds of well-defined procedures, for both normal andemergency conditions, by January 1, 1980. NRC needs
to develop new criteria for determining acceptable
technical qualifications to design and constructnuclear power plants.

*(4) IT IS IMPORTANT TO ATTRACT HIGHLYQUALIFIED CANDIDATES FOR THE POSITIONSOF SENIOR OPERATOR AND OPERATOR SUPERVISOR. Pay scales should be high enough to attractsuch candidates.

*

*

*

(4) NRC has taken actions and will do more to substantially increase the qualifications of operating plant
personnel (see next heading). NRC agrees it will benecessary for utilities to increase their pay scales.

*(5) SUBSTANTIALLY MORE ATTENTION AND
CARE MUST BE DEVOTED TO THE WRITING,
REVIEWING, AND MONITORING OF PLANT
PROCEDURES. Clearer wording, sound and practical content, clear diagnostic instructions forJdentifying abnormal occurrences, and insistence on the partof utility and vendor management on the early cureof safety questions (with deadlines, sanctions for delays, dissemination of results) are all recommended.

*(5) NRC believes that licensees must evaluate and incorporate operating experience into their procedures,
has ordered detailed analyses of small break loss-ofcoolant accidents for all B&W operating reactors, andhas ordered new analyses and procedures by all operating reactor licensees for responding to off-normal
events which can be aggravated by operator action.
Procedures which assist the operator in responding
to inadequate core cooling have also been prescribed.
Studies of the effects of stress on operator actions areunderway and human factors will be afforded a prominence equal to that given equipment in NRC systems
safety evaluations.

* **(6) STATE RATE-MAKING AGENCIES SHOULD
GIVE EXPLICIT ATTENTION TO THE SAFETYIMPLICATIONS OF RATE-MAKING WHEN THEYCONSIDER COSTS ON "SAFETY-RELATED"CHANGES.

**(6) NRC agrees and will consider further its role in the
resolution of the problem and examine whether otherfinancial considerations, such as deadlines for ratemaking purposes or tax exemptions, affect the safetyof a nuclear power plant.

**50==================================================PRESIDENT'S COMMISSION (PC) RECOMMENDATIONSON THE TRAINING OF OPERATING PERSONNELPC RECOMMENDATIONSNRC RESPONSES(1) AGENCY-ACCREDITED TRAINING INSTITUTIONS FOR OPERATORS AND SUPERVISORSSHOULD BE ESTABLISHED. Highly qualified instructors, high standards, and an emphasis on fundamentals of nuclear power plants and possible healtheffects thereof are recommended, and the training ofoperators to respond to emergencies. The institutions
could be national, regional, or specific to nuclear
steam systems; reactor operators should be required tograduate from one of them, with exemptions only
when there is documented evidence that the candidate
has equivalent training; the institutions should be subject to periodic reaccreditation by NRC; candidates
must meet entrance requirements.

(1) Although it agrees with the objective underlying
the recommendation, NRC is not convinced that ac~creditation by NRC is the best way to proceed (although it does not object, in the long term, to having
operators trained in a few, high-quality, accredited
institutions closely controlled by NRC). But NRC's
approach to date has aimed at upgrading the training
requirements while leaving the choice of where to conduct training to the utility. The Institute for Nuclear
Plant Operations established by the industry intends
to give training to utility management and to instructors involved in operator training, and if the Institutecan become the accrediting authority for reactor operator training, it might be preferable, although NRC
will certainly be more deeply involved in auditing andmonitoring training than ever before.

If<

If<

If<

(2) INDIVIDUAL UTILITIES SHOULD BE RESPONSIBLE FOR TRAINING OPERATORS WHO
ARE GRADUATES OF ACCREDITED INSTITUTIONS IN THE SPECIFICS OF A PARTICULAR
PLANT. The operators should be examined and licensed by the NRC both at initial licensing and at
relicensing; operators must pass every portion of theexamination, and supervisors of operators should
have, at a minimum, the same training as operators.

'"

'"

If<

If<

If<

'"

(2) Utilities are now responsible for training operatorsin the specifics of a particular plant. Operators are
initially examined and licensed by NRC, but licenses
are renewed every 2 years afterward without NRC
examination. NRC is taking action to reexamine operators for license renewal, to increase the overall passing grade and require it for each portion of the test
(effective now), and will continue to require supervisors to have at least the same training as operatorsand be licensed as senior operators, as before.
Managers at certain levels may also be required to belicensed as senior operators.
If<

'"

If<

J

(3) COMPREHENSIVE ONGOING TRAINING
MUST BE GIVEN TO MAINTAIN OPERATORS'LEVEL OF KNOWLEDGE. The training must becontinuously integrated with operating experience,with emphasis on diagnosing and controlling complex
transients, and on fundamental understanding of re"actor safety. Each utility should have ready access toa control room simulator, and operators and supervisors should be required to train regularly on it.
Retention of operator licenses should be made contingent upon simulator performance.
If<

'"

'"

(3) NRC requires ongoing training and requalificationof operators with annual examinations conducted by
the utility. Requalification programs are being revised
to give more emphasis to diagnosing and controllingcomplex transients, improving the fundamental graspof reactor safety, and taking account of operating
experience. In the future, NRC will administer requalification exams. The use of simulators will be
required in operator training and retraining and forrecertification. NRC is considering a requirement thatutilities upgrade training for all plant personnel, over
and above the recommendation cited.
If<

If<

=================================================51
PC RECOMMENDATIONSNRC RESPONSES(4) RESEARCH AND DEVELOPMENT SHOULD
BE CARRIED OUT ON IMPROVING SIMULATION
AND SIMULATION SYSTEMS, to bring a higher
level of realism to operator training, including simulated transients, and to improve diagnostics and general knowledge of nuclear plant systems.

(4) NRC believes that different types of simulators
are needed to upgrade training, on the one hand, andrefine diagnostic techniques, on the other. Explicit requirements are being readied for the simulator exercises to be included in operator training, coveringnormal and abnormal situations and response to multiple and concurrent failures. NRC will undertake extensive research in this area.

*

*

*

*

PRESIDENT'S COMMISSION (PC) RECOMMENDATIONSBASED ON ITS TECHNICAL ASSESSMENT
PC RECOMMENDATIONSNRC RESPONSES(1) EQUIPMENT SHOULD BE EVALUATED AC.
CORDING TO THE EXTENT TO WHICH IT INFORMS AND ASSISTS OPERATORS TO HELP
THEM PREVENT ACCIDENTS AND DEAL WITH
THOSE THAT DO OCCUR. Instruments should give
both monitory and precursory information, e. g., indications of the full range of temperatures in the reactor under normal or abnormal conditions, and indication of the actual position of valves. Computer
technology should be used to furnish clear displays tooperators and supervisors of measurements relevant toaccident conditions and advance warnings of developing conditions. In the interim, for TMI and similar
plants, grouping of key measurements should be considered, with distinct warning signals on a single
panel available to a specific operator and a duplicate
panel to the supervisor.

(1) NRC agrees with all PC recommendations on improved control room designs and believes that theneed for improved design is one of the most importantof TMI lessons. Actions have been taken to improve
the ability of operators to prevent or cope with accidents by improving the information available to them.
Revised procedures and operator training in recognizing inadequate core cooling are ,required to be completed by the end of 1979 at all operating reactors.
Instrumentation to monitor water level in the reactor
and pressure, water level, radiation and hydrogen in
the containment will be required by the end of 1980,
as will other safety items designed to inform the operators clearly and fully. The most important new requirement is the year-long review of control rooms
employing experts in human factors and personequipment interaction. In the long term, NRC is encouraging completion of an industry standard on control room design and will carry out research in this entire area.

* *(2) EQUIPMENT DESIGN AND MAINTENANCE
INADEQUACIES AT TMI SHOULD BE STUDIED
WITH A VIEW TO MITIGATING THE CONSEQUENCES OF ANY SIMILAR FUTURE OCCURRENCE. Iodine filters, the hydrogen recombiner, thevent gas system, containmnet isolation, reportingof water and radiation levels in containment, and thefast analysis of containment samples all merit review
and correction.
* * *(3) MONITORING INSTRUMENTS AND
RECORDING EQUIPMENT SHOOLD BE PROVIDED TO RECORD CONTINUOUSLY ALL CRITICAL PLANT MEASUREMENTS AND CONDITIONS.

*

*

*

*
(2) The NRC staff has required all licensees to fix six ofthe seven types of components cited by January 1,
1981. Iodine filtration is the subject of ongoing study
and criteria development which includes other postaccident radiation control and treatment matters.
Requirements for design changes redressing otherequipment and maintenance deficiencies have also
been imposed.

*(3) NRC is in complete accord. General criteria forsuch a requirement were developed by the Lessons
Learned Task Force in the form of instrument
readings which characterize the plant's safety status.
NRC has required that recording equipment and instrumentation be present in the new on-site technicalsupport centers by January 1, 1981.
*

52

PC RECOMMENDATIONSNRC RESPONSES(4) CONTINUING IN-DEPTH STUDIESSHOULD BE INITIATED ON THE PROBABILITIES AND CONSEQUENCES OF NUCLEAR
POWER PLANT ACCIDENTS, including the consequences of meltdown. The studies should cover both
onsite and offsite effects and encompass a varietyof small break loss-of-coolant and multiple failure
accidents, with particular attention to human failures.
Such studies should be useful in planning for recovery and cleanup after a major accident and in
modifying plant design to help prevent or mitigate
accidents (e. g., venting hydrogen from the reactor
coolant system); they could be carried out by
industry or other organizations under NRC or other
Federal sponsorship.

(4) NRC agrees and has increased or redirected
its current program, requiring licensees to analyze
small break loss-of-coolant accidents assuming multiple equipment failures. These are complete andrevisions of procedures and training have beeneffected. Crystal River Unit 3, a B&W operating
plant, is included in the Integrated Reliability Evaluation Program, as well other operating plants andpossibly new operating plant licensees. NRC is also
redirecting its research program to take in more probable transients and small break accidents, and is investigating core melt phenomena, including datafrom TMI relevant to recovery and cleanup after
a major accident. Some specific deficiencies revealed
at TMI and present elsewhere will be, as recommended, corrected before the end of 1980, but NRC believes
that, since the deficiencies existed because this kind of
TMI accident had not been considered in design andevaluation of the plant, mitigatory design features addressed to core damage and core melting may be required.

(5) STUDY SHOULD BE MADE OF THE CHEMICAL BEHAVIOR AND THE RETENTION OF
RADIOACTIVE IODINE IN WATER, which resuI ted in the very low release of radioiodine to theatmosphere in the TMI accident. The information
should be taken into account in the studies of the consequences of other small break accidents.

(5) NRC agrees that more information is needed on
the realistic behavior of iodine, other radioisotopes
and chemicals in the primary coolant systems ofseverely damaged reactors, and will conduct thenecessary research.

(6) BECAUSE OF HEALTH HAZARDS ASSOCIATED WITH THE CLEANUP AND DISPOSALPROCESS, CLOSE MONITORING OF THE
CLEANUP PROCESS AT TMI AND OF THE
TRANSPORTATION AND DISPOSAL OF THE
RADIOACTIVE MATERIAL THERE IS RECOMMENDED. As much data as possible should be preserved and recorded about the conditions within thecontainment building for future safety analyses.

(6) NRC agrees and has had a continuing presence atthe site to monitor, audit and review the cleanup
underway. As much important data as possible will bepreserved and recorded for future use. NRC has alsodecided to prepare a programmatic environmental
impact statement on the decontamination and disposalof wastes from the TMI accident.

.

(7) AS PART OF THE NORMAL SAFETY ASSURANCE PROGRAM, EVERY ACCIDENT OR NEWABNORMAL EVENT SHOULD BE SCREENEDTO ASSESS ITS IMPLICATIONS for the existing
system design, computer models of the system, equipment design and quality, operations, operator training, training simulators, plant procedures, safetysystems, emergency measures, management andregulatory requirements.

*

•

.

'"

(7) NRC agrees on the need for thorough investigationof accidents and abnormal events and believes thatthe initiatives on operating experience evaluation, inclose coordination with inspection and enforcement
activities for the especially significant events, willmeet the intent of this recommendation.

•

===================================================53
PRESIDENT'S COMMISSION (PC) RECOMMENDATIONSON WORKER AND PUBLIC HEALTH AND SAFETY
PC RECOMMENDATIONSNRC RESPONSES(1) EXPANDED AND BETTER COORDINATEDRESEARCH INTO HEALTH-RELATED RADIATION EFFECTS SHOULD BE ESTABLISHED, andshould include, among others, study of the biologicaleffects of low levels of ionizing radiation; acceptablelevels of ionizing radiation to which the general public
and workers may be exposed; means for mitigating theadverse health effects of exposure to ionizing radiation; and the genetic or environmental factors which
predispose individuals to incurring adverse effects.
The research should be coordinated with the National
Institutes of Health and other Federal agencies.

(1) NRC agrees with the recommendation. During
1978-79, the NRC staff worked in an interagency
project chaired by the Department of Health, Education and Welfare, which also concluded that there wasneed for this kind of research. Thus, the interagency
committee on radiation research, chaired by the National Institutes of Health, was established in early
1979, with NRC as a member. Topics cited by the PC
will be introduced by NRC as agenda items for action
by the committee.

*

*

*

(2) NRC POLICY STATEMENTS OR REGULATIONS CONCERNING RADIATION-RELATEDHEALTH EFFECTS, INCLUDING REACTOR SITING ISSUES, SHOULD BE SUBJECT TO REVIEWAND COMMENT BY THE SECRETARY OF THE
DEPARTMENT OF HEALTH AND HUMAN SERVICES. A time limit should be placed on such review toassure expeditious treatment.
*(2) NRC agrees with the value of Federal oversight of
NRC activities that affect public health. But NRC
believes that a more effective and balanced result
would be achieved through the role envisioned for theFederal Radiation Policy Council that the President
has decided to establish.

*(3) AN INCREASED PROGRAM, AS A STATE AND
LOCAL RESPONSIBILITY, FOR EDUCATINGHEAL TH PROFESSIONALS AND EMERGENCYRESPONSE PERSONNEL IN THE VICINITY OF
NUCLEAR POWER PLANTS SHOULD BE
CREATED.

**(3) NRC agrees with this recommendation and, although the suggestion is for a State and local program,
NRC intends to give guidance and help in meeting
their needs. In particular, NRC will supplement NRCIEPA guidance already available to States on the preparation of emergency response plans to provide moredetailed guidance on the education and training of
personnel who will respond to emergencies at nuclear
power plants. In addition, NRC has offered and willcontinue to offer technical assistance to the States in
the preparation or upgrading of emergency responseplans.

*(4) UTILITIES MUST MAKE SUFFICIENT ADVANCE PREPARATION FOR THE MITIGATIONOF EMERGENCIES, by having radiation monitors
available for normal or off-normal conditions; byhaving the emergency control center for health physics
operations and analytic laboratory in a well-shieldedarea with its own air supply; by having enough instru-(4) The recommendation of the NRC Task Force on
Emergency Preparedness to expand coverage andimprove offsite monitoring capability for accidents isbeing implemented by all operating plant licensees,and NRC has increased its capability in this area.
Requirements for onsite monitoring for accident diagnostics and health physics purposes recommended by

54

PC RECOMMENDATIONSNRC RESPONSES

ments, respirators, and other equipment for normal or
off-normal conditions; and by performing adequate
maintenance on all such health-related equipment.

the Lessons Learned Task Force are also being implemented. Requirements for emergency health physics
control centers and health physics equipment are being upgraded. These actions should substantially improve utility capability.

II<

II<

II<

(5) AN ADEQUATE SUPPLY OF POTASSIUM
IODIDE FOR PROTECTION AGAINST RADIATION EFFECTS ON THE THYROID SHOULD BE
AV AILABLE REGIONALLY FOR DISTRIBUTION
TO THE GENERAL POPULATION AND WORKERS AFFECTED BY A RADIOLOGICAL EMERGENCY.
II<

II<

II<

II<

II<

(5) NRC agrees and will require licensees to have adequate supplies of this agent available for nuclear
power plant workers. For the general population,
NRC expects to make its availability a necessary partof an acceptable State emergency response plan. Plans
are not complete as to how and how much of the agent
should be stockpiled and distributed; studies areunderway.
II<

II<

PRESIDENT'S COMMISSION (PC) RECOMMENDATIONSON EMERGENCY PLANNING AND RESPONSE
PC RECOMMENDATIONSNRC RESPONSES(1) EMERGENCY PLANS MUST DETAIL CLEARLY AND CONSISTENTLY THE ACTIONS PUBLIC
OFFICIALS AND UTILITIES SHOULD TAKE
WHEN OFFSITE RADIATION DOSES OCCUR.
The State within which a prospective nuclear power
plant will be sited should have an emergency response
plan reviewed and approved by the Federal Emergency Management Agency (FEMA) before an
operating license is granted. FEMA should have theFederal responsibility for radiological emergency
planning and should consult with other agencies, including the NRC and health and environmental agencies. The State should coordinate its planning withthe utility and local officials, and States with plants
now operating should upgrade, without delay, theirplans to conform with FEMA requirements.

(1) NRC agrees with the substance of the recommendation and has moved to upgrade plans in Stateswith operating plants. Rulemaking has been initiatedto raise emergency preparedness standards and an
extensive review of all aspects of response capability
is underway. A joint letter has been issued by FEMA
and NRC confirming the former's lead role in Federal
emergency planning and declaring joint responsibility
for concurring in State emergency response plans priorto NRC's issuance of an operating license. NRC is considering a rule that would make such issuance contingent upon approval of State plans within a fixed
time frame.

II<

II<

II<

(2) PLANS FOR PROTECTING THE PUBLIC
FROM OFFSITE RADIATION RELEASESSHOULD BE BASED ON TECHNICAL ASSESSMENT OF VARIOUS CLASSES OF ACCIDENTS
THAT CAN TAKE PLACE AT A GIVEN PLANT. No
single plan based on fixed distances and responses cansuffice; planning should involve the identification ofseveral different kinds of accidents with different radi-

II<

II<

II<

(2) The basis for emergency response planning has
been under examination at NRC for some time. An
NRC/EPA task force published the results of an extensive study in December 1978 and its conclusions wereconsistent with this recommendation. In October
1979, the NRC Commissioners endorsed the concept ofa flexible planning base, including emergency planning over much larger areas than before. The base re-================================================55
PC RECOMMENDATIONSNRC RESPONSES

ation effects. For each kind there should be clear
criteria for the appropriate response at various distances, such as instructing people to remain indoors
for a time, providing special medication, or ordering
an evacuation. Response plans should be keyed tovarious possible scenarios and activated when thenature of the potential hazard is clear. Plans should
exist for protecting the public from radiation levels
lower than those in current NRC-prescribed plans.
And all local communities should have funds and tech.
nical support adequate for preparing the plans
recommended.

quires that specific scenarios be used to test the adequacy of plans and that the activation of emergency
response be keyed to various plant conditions according to revised emergency action guidelines publishedin September 1979. NRC currently uses the EPA
protective action guides, but will give greater emphasis in the new action level guidance on the potential for exposure as distinct from the actual exposurelevels. An NRC staff study on funding problems ofState and local governments was recently publishedand is under consideration by NRC; it discusses theneed for and possible sources of such funding.

*

*

*

(3) RESEARCH SHOULD BE EXPANDED ON
MEDICAL MEANS FOR PROTECTING THE PUBLIC AGAINST VARIOUS LEVELS AND TYPES OFRADIATION. This research should include exploration of appropriate medications that can protect
against or counteract radiation.
***(4) IF EMERGENCY PLANNING AND RESPONSE
TO A RADIATION-RELATED EMERGENCY IS TO
BE EFFECTIVE,- THE PUBLIC MUST BE BETTER INFORMED. A program is needed to educate
the public on how nuclear power plants operate, onradiation and its health effects, and on protective
actions required in an emergency.

*

*

*

(5) COMMISSION STUDIES SUGGEST THAT DECISION-MAKERS MAY HAVE OVERESTIMATED
THE HUMAN COSTS, IN INJURY AND LOSS OF
LIFE, IN MANY MASS EVACUATION SITUATIONS. Further study is needed into the human costsof mass evacuation and into the question of whether
radiation-related evacuations differ from those occasioned by other events. Such studies should take into
account the effects of improved emergency planning,
public awareness of the planning, and costs.
***(6) PLANS FOR PROVIDING FEDERAL TECHNICAL SUPPORT, SUCH AS RADIOLOGICALMONITORING, SHOULD CLEARLY SPECIFY
THE RESPONSIBILITIES OF THE VARIOUS SUPPORT AGENCIES AND THE PROCEDURES BY
WHICH THEY PROVIDE ASSISTANCE. Existingplans, expecially those of the Interagency Radiological
Assistance Plan and the various memoranda of understanding among the agencies, should be reexamined
and revised by Federal authorities in the light of TMI
and better coordination and more efficient Federal
su pport prOVided for.

*

*

*

(3) NRC agrees that such research is needed and willencourage the Department of Health and Human
Services to take steps in this area.

*

*

*

(4) NRC agrees but believes that a broad public information program would be more appropriately handled by other agencies. Better information on radiation
risks is among the subjects to be addressed by theplanned Federal Radiation Policy Council. NRC
will require, however, that licensees keep the public
informed on a continuing basis of the nature ofhazards in a radiation emergency and of actions thatmight have to be taken. Periodic response drills on thepart of local and State organizations should contributeto this awareness.
**(5) NRC agrees th~t further study should be done on
this and other protective actions.

*(6) NRC agrees that improvements are needed andhas efforts underway to reexamine and revise Federal
interagency agreements on emergency assistance.

56

PRESIDENT'S COMMISSION (PC) RECOMMENDATIONSON THE PUBLIC'S RIGHT TO INFORMATIONPC RECOMMENDATIONSNRC RESPONSES(1) FEDERAL AND STATE AGENCIES, AS WELL
AS THE UTILITY, SHOULD MAKE ADEQUATEPREPARATION FOR A SYSTEMATIC PUBLIC INFORMATION PROGRAM, so that when a radiation
emergency occurs, they can provide timely and accurate information to the news media and the public in
a form that is understandable. Assignments of briefing
responsibility and availability of informed sources arenecessary to reduce confusion and inaccuracy. The
utility has primary responsibility for providing information on the status of the plant to the news media
and the public, as it has for the management of theaccident. The NRC should also be available to provide
background information and technical briefings. A
designated State agency should convey all information
related to State decisions on protective actions (including evacuations) and to offsite radiation releases.
This agency should set up the means to keep local officials informed and to coordinate briefings to discuss
Federal involvement in any evacuation measures.

(1) The procedure used before TMI was that NRC
public affairs staff would be sent to an accident site tosupport NRC personnel in communicating with themedia, but not to take charge of information activities. At TMI, the NRC in fact took over public information responsibilities on March 31. Although thisrecommendation prescribes a background role for
NRC, it seems more realistic that the Federal regulator be in a position to talk about an emergency situaation, since NRC would expect the State and the
public to look to NRC for authoritative information on
the situation. NRC believes it would be more effective to have Federal, State, and utility personnel
operate out of a single press center and, wheneverpossible, give a unified view of the situation.

(2) BECAUSE THE OFFICIAL SOURCES OF INFORMATION MUST MEET THE NEEDS OF THE
MEDIA FOR INFORMATION WITHOUT COMPROMISING THE EFFORT OF OPERATIONALPERSONNEL TO MANAGE THE ACCIDENT, it
is recommended that those who brief the news media
have direct access to informed sources of information,that technical liaison people be designated as contacts
for the briefers and the media, and that primary official news sources have plans for promptly setting uppress centers fairly close to the site, properly equippedand staffed.

(2) NRC agrees with the recommendations and willconsider requirements to assure that licensee plans willachieve them. Licensees are now required to identify
offsite emergency control centers where the utility,
Federal, State, and local officials can gather. A press
center would be established either at the off-site
emergency control center or nearby, which will facilitate State activities set forth in the preceding
recommendation.

(3) SPECIAL RESPONSIBILITIES ON THE NEWSMEDIA TO PROVIDE ACCURATE AND TIMELYINFORMATION REQUIRE THAT all major media
hire and train specialists familiar with reactors andradiological language, and all other media in thearea of nuclear power plants should have plans forsecuring such services in an emergency; reporters try
to place complex information in an understandable
context and allow the public to decide the hazard totheir health and safety; reporters try to avoid raising"what if" questions needlessly and try to understandexpressions of uncertainty and probability from thesources of information.

(3) NRC agrees and will urge the professional societies, such as the American Nuclear Society or the
Health Physics Society, to sponsor seminars for thenews media where reporters can learn how nuclear
power plants operate and about radiation effects.
NRC will consider in ongoing rulemaking whether the
training program required to be developed by thelicensee for local officials could be extended to include
local news media personnel.

57

PC RECOMMENDATIONSNRC RESPONSES(4) STATE EMERGENCY PLANS SHOULD INCLUDE PROVISION FOR CREATION OF LOCALBROADCAST MEDIA NETWORKS FOR EMERGENCIES THAT WILL SUPPLY TIMELY AND ACCURA TE INFORMATION. Arrangements should bemade to have knowledgeable people available to go on
the air and clear up rumors and explain conditions.
Communications between State officials, the utility,
and the network should be prearranged to handle thepossibility of an evacuation announcement.

(4) NRC agrees the proposal has merit and will incorporate recommendations accordingly in guidance to
the States. It will also consider in the ongoing rulemaking on emergency preparedness whether there is an0ed to include requirements for licensee planning andcoordination to disseminate information to the public
on these local broadcast networks and to provide information to such networks in the event of an accident.

*

*

*

*

*

*

(5) THE PUBLIC IN THE VICINITY OF A NUCLEAR POWER PLANT SHOULD BE ROUTINELYINFORMED OF LOCAL RADIATION MEASUREMENTS THAT DEPART APPRECIABLY FROMNORMAL BACKGROUND RADIATION, whether
from normal or abnormal operation of the nuclear
power plant, from a radioactivity cleanup operationsuch as that at TMI, or from other sources.

(5) NRC agrees with this recommendation, which isconsistent with its current practice, in which publicannouncements are made on any releases to the environment from licensed facilities that appreciably
exceed NRC limits (which are small in comparison
with normal background, but are in addition to normal background). Most licensees also issue such announcements.

In addition to providing the Executive Office of
the President with responses to each of the President's
Commission's recommendations, the NRC citedseveral examples of considerations and actions it had
taken as a result of TMI which were outside the scopeof the PC recommendations. Seven such examples
were given.

regulatory safety goal should comprise both evaluativeand quantified reliability criteria, applicable to thedevelopment of any new regulatory requirements andto a decision on backfitting requirements to existing
plants.

(1) Generic Requirements for Design Features forCore Melt Consequence Mitigation. Severe coredamage did occur at TMI, but significant exposure of
the public was prevented because radiation releases
were, for the most part, successfully kept in the containment bUilding. There is substantial evidence thatthe residual risks of core melt accidents can besignificantly reduced if some of the potential modes ofcontainment failure can be prevented or controlled.
The NRC Lessons Learned Task Force has recommended that this issue-whether to require additional
design features and training for core melt accidents-be revised through the rulemaking process.

(2) Expanded Reactor Safety Goals, Including
Quantification of Reliability. The President's Commission endorsed the conservative use of safety-costtradeoffs, but did not confront the fundamental question as to just what level of safety is desired and acceptable. The Advisory Committee on Reactor Safeguards
and the NRC Lessons Learned Task Force have recommended that policy guidance be developed within
NRC on what is an acceptable safety goal of reactor
regulation,reflecting a synthesis of views and priorities
and setting forth an objective sufficiently clear for thestaff to employ in day-to-day decisionmaking. This

(3) NRC Emergency Response Capabilities. Eventsat TMI demonstrate that NRC has an important rolein auditing and monitoring the licensee's actions, and
NRC is strengthening the crisis management andtechnical capabilities of its emergency management
staff. The emergency response teams of the NRC Office of Inspection and Enforcement are being tested
and actually dispatched to various sites. NRC is also
specifying the content and transmission requirements
for a nuclear data link from all operating plants to its
Operations Center.
(4) Compensating Features for Plants with High
Population Density Sites. NRC is considering the needfor additional protective action-such as shutdown,
reduced power or additional design features-for currently operating plants located in densely populatedareas.

(5) Licensing of Operations Personnel in Addition
to Reactor Operators and Their Supervisors. NRC isconsidering making it a requirement that certain
nuclear power plant personlJ.el other than reactor
operators and supervisors be licensed. TMI indicatedin various ways that plant safety can be affected bypersons in many positions, including managers,
engineers, auxiliary operators, maintenance personnel, and technicians. The Institute of Nuclear Power58

Operations, recently established by the nuclear industry, may have a role to play in this area.
(6) Plant Security During an Emergency. A needfor clear instructions for plant security during an
emergency was brought home by TMI, particularly toensure that access control measures remain effective
but do not hamper recovery operations.
(7) Worker Protection. Significant deficiencies in
the worker protection program at TMI have been
disclosed and, concerned that the problems may be
widespread, NRC is developing new generic requirements in this area.

Two of the five NRC Commissioners made separate
supplemental responses to the President's Commission
findings and recommendations. These are summarized
below.
Commissioner Bradford's Views. Commissioner
Peter A. Bradford expressed his judgment that, while
the PC report was helpful and insightful in a numberof areas (including recommendations on the NRC, on
operating personnel training, technical assessment,
and emergency planning), it was "a flawed document"in three respects. First, the major recommendation fora restructuring of the nuclear regulatory process "doesnot make good sense." Second, there are a number ofareas to which the report could have spoken but did
not. Third, there is "no clear relationship between the
narrative, the findings, and the recommendations,with the result that some important findings do not
result in recommendations while some of the recommendations find little support elsewhere in thereport."On the first flaw, the Commissioner felt that theconcept of an independent agency headed by a "single
administrator appointed by the President ... to serveat the pleasure of the President" presented a "contradiction in terms," since an agency cannot be independent if its head is removable at the pleasure of
the President. Further, the "more this point is corrected by the granting of true independence to theagency the more undesirable it will be to vest what
will become quite sweeping powers in a single individual."
The problems within NRC to which the recommendation is addressed are of two kinds: an "attitudinal"
problem, which shows up in the agency's failure to
pursue the questions which would have led it to
discover the vulnerabilities now revealed by TMI; andthe diversity of views among the NRC Commissioners
which may make it difficult for the agency to correct
itself. While the Commissioner agreed that the second
problem was curable by setting up a single administrator, as recommended by the President's Commission, "it is also curable through changes within thecurrent Commission structure" which would constitute a "potentially faster and certainly wiser" courseof action. The Commissioner pointed out that the only
real benefit of the single-administrator proposal (orproposals to reinforce the authority of the Chairman
or the Executive Director for Operations) is "that itprovides a shortcut away from the perceived stalemate
at the current Commission." He felt that these proposals "ignore the fact that collegial agencies are
perfectly capable of moving rapidly and innovatively
in new directions as long as they have a coherent andpredictable majority that includes the Chairman andthat supports the chief operational officers."
A number of items were cited by the Commissioner
on which he believed the President's Commission"could usefully have taken a position had time permitted."• On the question of whether and when evacuation
was warranted at TMI, he notes that the PC
report "said nothing about the validity of the actual recommendation that was made, This seems
to me to be an oversight of some magnitude, forsuch decisions are often likely to involve theallocation of unquantifiable uncertainties. It
would be very useful to know whether thesetwelve citizens, .. feel that a greater or lesser setof evacuation advisories were in order at different
times during the accident."• The report does not discuss "the pros and cons ofintervenor funding .. , an essential tool to enable
the proposed Public Counsel to guarantee effective outside skeptical participation in the licensing process."• The PC report is "blurred as to what the fundamental standard for the safety of nuclear powershould be . . . . [T]he considered view of twelve
laymen on this subject would have been extremely
valuable. Instead, one finds statements to the effect that 'accidents as serious as TMI should not
be allowed to occur in the future.' . . . [S]ome
statement as to how this group regarded the acceptability of risks from nuclear power plants inthe context of other technologically imposed risks
would have been a helpful guidance." The NRC isgoing to have to "fill the void with a rulemaking."• There is no acknowledgment in the PC report of"the strides already achieved since Three Mile
Island by the Nuclear Regulatory Commission.
. , . This oversight would be easier to understand
if it were explicitly acknowledged and explained.
It would also be easier to understand if the TMICommission had not gone out of its way to pat thenuclear industry on the back for having recently
created the Institute of Nuclear Power Operations,"• The report speaks repeatedly of examples of AEC
promotional attitudes and practices within the
NRC but gives no specifics, The statements "tendto tar everyone with the same brush, and they are

=====================================================59
not helpful in setting a clear course of corrective
action."• While the report criticizes the NRC's "single
failure criterion," it makes no specific recommendation on the subject. If the criterion is to beabandoned, the implications for the nuclear
licensing process "are considerable and would
almost certainly result in extensive redesigningand backfitting to plants already under construction or in operation." If this is the recommendation of the report, it should have been made explicit.
• The PC report "lays a gentle and indecipherable
hand on the state ratesetting process." In the relation between financing and safety, there are "atleast two areas of much greater significance ...
the timing of state decisions that create an incentive to rush a plant into service (this allegation
was specifically made in regard to TMI) and theInternal Revenue Service's practice of assuming
for tax purposes that the plant was in service forthe full calendar year if it is ill commercial operation by midnight on December 31." Both questions are under study by NRC and "it might bewell to ask the Internal Revenue Service and the
National Association of Regulatory Utility Commissioners to have a look at them as well."• On the subject of worker and public health andsafety, the report "contains nothing on the vitalsubject of making sure that workers are adequately informed and trained with regard to radiationand its hazards. It also says nothing about theneed to assure that workers who raise safety- orradiation-related concerns are adequately protected against reprisals by their management."• The report fails to note that the Atomic Energy
Act "currently preempts the States from setting
most radiological health and safety standards involving nuclear power plants .... [I]f the states
had a role in this area, they would no longer findthemselves excluded from nuclear power plantradiation regulatory matters until the moment at
which something really goes wrong and they are
expected to step in and cope effectively with theoffsite consequences."• The report "says nothing about the effect of theattitudes of the Congressional Oversight Committees on the quality of the nuclear regulatory process:' The approach of the former Atomic Energy
Commission cited so often and so critically by the
President's Commission "was shaped by thedemands that were laid on the AEC by the Joint
Committee on Atomic Energy. Anyone trying tounderstand where nuclear regulation went astraymust realize that the AEC was responding notsolely to its own or to Executive Branch notions ofdesirable Atomic Energy policy, but also to thecontinuing pressure for results from the one congressional committee to which it was answerable.
The relationship as I understand it was a mutually reenforcing one, but the continuing role of theCongress setting the tone for nuclear regulation
should not be overlooked:'
Commissioner Gilinsky's Views. Commissioner Victor Gilinsky also put on record certain personal viewson the report of the President's Commission. On thebasic finding of a need for fundamental change, theCommissioner was in agreement, noting that publication of the report and the attention it received,
especially from the President, strengthens the ha.nd of"those concerned with improving nuclear safety andfurther shifts the burden of proof to those who would
do less rather than more." The Commissioner expressed agreement with "almost all" of the findings andrecommendations of the report, but stated, "I feelcompelled to add that when we get below the generallevel, down to the nitty-gritty of reactor regulation,
the report is less helpful."
The inventory of items that need fixing-operator
training, emergency planning, improved use ofoperating information, etc.-are "almost all .. , thesubjects of major NRC actions which were initiatedbefore the report's publication:' The more difficult
questions "in each case are: What precisely needs to be
done? Are NRC actions sufficient?" The President's
Commission decided that the present NRC is unable tofulfill its responsibility for providing an acceptablelevel of safety, but the PC report "is silent on what an
acceptable level is." It is up to the NRC, the Commissioner concludes, to "get more specific about overall
standards for nuclear safety-on what is safe enough."
The section of the PC report dealing with utility
management deficiencies carries "no attempt to judge
whether these deficiencies are characteristic of the industry. Without such a determination, it is impossible
to judge the overall system for public protection."
The report also fails to deal with the adequacy of the
TMI licensee's communication to government
authorities of plant conditions-high core temperatures and the containment hydrogen explosion-on thefirst day of the accident. "I regard this as a vital question," the Commissioner declared. "Given the dangers
inherent in nuclear plants we have to be confident thatthe utilities will report promptly any conditions that
require public 'protection."
The report "never comes to grips with the questionof whether an evacuation should or should not havebeen ordered," a decision which "is critical to forming
a judgment on the [Nuclear Regulatory] Commission's
responses and to planning further response,"
On the subject of NRC Commissioners' isolationfrom the licensing process, the Commissioner suggeststhat the single administrator called for in the PC
report "would be even more removed from the licensing proceedings" because, as the report proposes it, the60President and Mrs. Carter toured the TMI site on Sunday, Aprill,
1979, and are seen above in the TMI-2 control room. At left is
NRC's Director of Reactor Regulation, Harold Denton, who wasdesignated the President's personal representative at the site for theduration of the accident.

appeal board decisions would not be reviewable by theadministrator. The Commissioner indicates that theexperience of NRC is that "leaving all appeals to the
Appeal Board leads to loss of policy control over the
licensing process." He urges that "[t]he Commissioners
need to be more involved in the adjudicatory reviews
rather than less."

The President's Response

The PC report recommends, "after seemingly
streamlining the NRC for emergencies by shifting to asingle administrator," that the NRC "stay out of dealing with emergencies altogether" and leave emergency
planning to FEMA and the handling of any accident-and public information related thereto-to
the utility. The Commissioner does not think it "wiseor realistic to downplay the NRC role to this extent."
The Commissioner also observes that the report, byemphasizing the human failures and "thereby vindicating the equipment," does not stress enough thatthe equipment "could have been designed to avoid thiskind of trouble."

On October 30, 1979, the President's Commission
on the Accident at Three Mile Island presented its final
report to the President. Following a period of study bya panel appointed from his staff, the President issued
his response to the recommendations of the PC reporton December 7, 1979. (The President's statement isreprinted on page 62 in its entirety.)
Among the salient points of the statement were the announcements that:• A reorganization plan for the NRC would be sent
to Congress in the next session which will
strengthen the role of the Chairman to lead theCommission in the development of a unified andmore reliable nuclear safety regulatory program.
• The President would appoint a new Chairman of
the NRC from outside the agency.
• A five-member expert advisory committee would
be established to monitor the progress of the61

NRC, other Federal agencies, the States, and theutilities in improving the safety of nuclear reac~
tors and in implementing recommendations of the
President's Commission. The committee would
report periodically to the President and the
public.
• The President was asking the NRC and ot}:teragencies to accelerate placement o~ a resident
Federal inspector at every reactor sIte and was
asking the NRC to evaluate the need for a Federal
presence in the control room of operating reactors.
• The President was directing that the Federal
Emergency Management Agency (FEMA) assume
responsibility for all offsite nuclear emergency
planning and response. A supplemental appropriation of $8.9 million would be submitted toCongress to enable FEMA to complete the reviewof State emergency plans in all States with
operating licenses by June 1980.
• The President was urging the industry to build on
the progress it had made since the TMI accident
to provide enhanced analysis and evaluation forsafety of the design, construction, and operationof plants and a greatly strengthened training,
retraining, and evaluation program for operatorsand supervisors. He asked the NRC to appraise
and reinforce these efforts.

• To assure that the lessons of TMI were expeditiously absorbed and applied, the President
was submitting a supplemental appropriation toCongress of $49.2 million for the NRC and $7
million for the DOE. These funds would allow
the collection and evaluation of data and speed
the implementation of reforms.
Affirming that he "agrees fully with the spirit andintent" of all recommendations of the PC report, thePresident chose to strengthen the NRC organization
through enhanced executive powers for the Chairman,
rather than by creation of a new agency. Since the collegial Commission, representing diverse and complementary views, would be retained, the President
chose not to create a I5-member oversight committee.
He did, however, announce his intention of
establishing a smaller advisory committee to report tohim on the progress of the NRC and others, as noted
above. The President urged the NRC to complete its
work of defining and effecting the reforms dictated byanalyses of TMI as quickly as possible and, in any
event, no later than May 1980. In doing so, the President observed that "we must resume the licensing pro~
cess promptly so that the new plants which we need toreduce our dependence on foreign oil can be built andoperated." He concluded by stating that "nuclear
power has a future in the United States-it is an optionthat we must keep open. I call on the utilities and their
suppliers, the NRC, the Executive Departments andagencies, and the State and local governments toassure that the future is a safe one."

62

Statement by President Carter on the
Kemeny Commission Report
I have reviewed the report of the Commission I established to investigate the accident at Three Mile Island nuclear power plant. The
Commission, chaired by Dr. John Kemeny, found very serious shortcomings in the way that both the government and the utility industry
regulate and manage nuclear power.
The steps I am taking today will help ensure that nuclear power
plants are operated safely. Safety has always been, and will remain,
my top priority.
As I have stated before, in this country, nuclear power is an energy
source of last resort. By this I meant that as we reach our goals forconservation, direct use of coal, development of solar power and syn. thetic fuels and enhanced production of American oil and natural, gas, we can minimize our reliance on nuclear power.
Many of our foreign allies must place greater reliance than do weon nuclear power, because thay do not have the vast natural
resources that give us many alternatives. We must get on with the jobof developing alternative energy sources-by passing the legislation I
proposed to the Congress, and by making an effort at every level ofsociety to conserve energy.
We cannot shut the door on nuclear energy.
The recent events in Iran have shown us the clear, stark dangers
that excessive dependence on imported oil holds for our Nation. We
must make every effort to lead this country to energy security.
Every domestic energy source, including nuclear power, is critical
if we are to free our country from its overdependence on unstablesources of high-priced foreign oil. We do not have the luxury of abandoning nuclear power or imposing a lengthy moratorium on its further use. A nuclear plant can displace up to 35,000 barrels per day.
We must take every possible step to increase the safety of nuclear
power production. I agree fully with the spirit and intent of theKemeny Commission's recommendations, some of which are within
my power to implement, others of which rely on the Nuclear
Regulatory Commission or the utility industry itself.
To get the government's own house in order I will take several
steps. First, I will send to Congress a reorganization plan tostrengthen the role of the Chairman of the NRC and provide this person with the power to act on a daily basis as the chief executive officer, with authority to put needed safety requirements and procedures in place. The Chairman must be able to select key personnel,
and act on behalf of the commission during an emergency.
Second, I will appoint a new Chairman of the NRC-someone
from outside that agency, in the spirit of the Kemeny Commission's
recommendation. In the meantime, I have asked Commissioner
Ahearne, now on the NRC, to serve as Chairman. Dr. Ahearne will
stress safety and the prompt implementation of the needed reforms.
In addition, I will establish an independent advisory committee tohelp keep me informed of the progress the NRC and the industry areachieving in making nuclear energy safer.
Third, I am directing the Federal Emergency Management Agency to head up all off.site emergency activities, and complete athorough review of emergency plans in all sta:tes with operating reactors by June.
Fourth, I have directed NRC and other agencies to accelerate ourprogram to place a resident federal inspector at every reactor site.

Fifth, I am asking all relevant government agenCies to implement
virtually all of the other recommendations of the Kemeny Commission.
A detailed fact sheet is being issued to the public, and a more extended briefing will be given to the press.
With clear leadership and improved organization, the Executive
branch and the NRC will be better able to act quickly on the criticalissues of improved training and standards, safety procedures, and theother Kemeny Commission recommendations.
But responsibility to make nuclear power safer does not stop withthe federal government. In fact, the primary day-to-day responsibility for safety rests with utility company management and suppliers of
nuclear equipment. There is no substitute for technically qualified
and committed people working on the construction, operation andinspection of nuclear power plants. Personal reponsibility must be
charged both at the corporate level and at the plant site. The industry
owes it to the American people to strengthen its commitment tosafety.
I calion the utilities to implement the following changes:
First, building on the steps already taken, the industry must
organize itself to develop enhanced standards for safe design, operation, and construction of plants.
Second, the nuclear industry must work together to develop and tomaintain in operation a comprehensive training, examination andevaluation program for operators and supervisors. This training program must pass muster with the NRC through accreditation of training programs.
Third, control rooms must be modernized, standardized andsimplified as much as possible to permit better informed decisionmaking during an emergency.
I challenge our utility companies to bend every effort to improve
the safety of nuclear power.
Finally, I would like to discuss how we manage the transition
period during which the Kemeny recommendations are being implemented. There are a number of new nuclear plants now awaiting
operating licenses or construction permits.
Licensing decisions rest with the NRC and, as the Kemeny Commission noted, it has the authority to proceed with licensing theseplants on a case-by-case basis, which may be used as circumstances
surrounding a plant dictate. The NRC has indicated, however, that
it will pause in issuing new licenses and construction permits in orderto devote its full attention to putting its house in order. I endorse theapproach the NRC has adopted, but I urge the NRC to complete its
work as quickly as possible, and in any event no latter than six
months from today.
Once we have instituted the necessary reforms to assure safety, wemust resume the licenSing process promptly so that the new plants
which we need to reduce our dependence on foreign oil can be built
and operated.
The steps I am announcing today will help assure our country of
the safety of nuclear plants. Nuclear power has a future in the UnitedStates-it is an option that we must keep open. I call on the utilities
and their suppliers, the NRC, the executive Departments and agencies, and the State and local governments to assure that the future is asafe one.

3
Reactor Regulation

In February 1980 NRC authorized fuel loading

and low-power operation of the TVA's
Sequoyali Plant.

The goals of the NRC in licensing and regulating
nuclear reactors in the United States are to protect the
public health and safety, to protect the quality of theenvironment, and to assure compliance with the antitrust laws in civilian nuclear activities. The reactor
licensing process is centered in the NRC Office of
Nuclear Reactor Regulation (NRR), where each proposed nuclear power plant is reviewed by a staff
drawn . from a broad spectrum of professional
disciplines. (See Appendix 1 for a description of the
NRR organization.)
The Three Mile Island accident revealed the need
. fot a number of changes in NRC's conception o( andapproach to nuclear safety, requiring action in theareas ?f human factors, operational safety, emergency
plannmg, nuclear power plant design and siting,health effects, and public information. Much of the~RR staff was involved in these efforts, participating
m the task forces on TMI Support, Bulletins andOrders, and Lessons Learned, as discussed in Chapter
2. This chapter deals with other matters related to thel~cens~ng and regulation of nuclear power plants: thehcensmg of reactor operators, unresolved safety issues
other t~chnical issues, advanced nuclear power plants:
protectIon of the environment, progress in standardizati,?n, antitrust and indemnity activities, and othersubjects.

Status of Nuclear Power Generation
As of September 30, 1979, there were 192 nuclear
power units either in operation, being built or being
planned, representing a total capacity of 187,000 net
megawatts electric (MWe). This total is 20 units fewer
than the total at the end of fiscal year 1978. Of the 192
units, 186 had entered the NRC licensing process, as
follows:• 70 licensed to operate, with a total capacity of
51,000 MWe.

• 91 with construction permits representing
100,000 MWe.
• 25 under review for construction permits, representing 29,000 MWe. (Initial construction work
was proceeding on four of these under limited
work authorizations.)
Of the remaining six units-those which had notentered the NRC licensing process-four had beenordered and two publicly announced.
Shortly after the close of fiscal year 1979, the Commission issued an "Interim Statement of Policy andProcedure," dated October 4, 1979, taking note of thevarious investigations-within and outside of
NRC-of the TMI accident still underway, and theimplications of those efforts for pending licensing actions. The statement affirmed that "new construction
permits, limited work authorizations, or operating
licenses for any nuclear power reactors shall be issued
only after action of the Commission itself .... In thesecircumstances no full adjudicatory decision whichauthorizes issuance of such a permit, authorization or
license shall be issued by an Atomic Safety and Licensing Board except after further order of the Commission itself."
On November 9, following publication of the reportof the President's Commission (see Chapter 2), Chairman Hendrie notified the Director of the Office of
Science and Technology Poli~y that new nuclear
power plants would not be licensed until NRC had
developed new or improved safety objectives and thecriteria by which to implement them.

Assistance From Other Agencies. Because of the necessity of reallocating a major portion of the staff of the Office of Nuclear Reactor
Regulation to the conduct and support of thenu,nerous . investigations and associated activities

64

THE LICENSING PROCESS
Obtaining an NRC construction permit-or a limited work
authorization, pending a decision on issuance of a construction permit-is the first objective of a utility or other company seeking tooperate a nuclear power reactor or other nuclear facility under NRC
license. The process is set in motion with the filing and acceptance of
the application, generally comprising ten or more large volumes ofmaterial covering both safety and environmental factors, in accordance with NRC requirements and guidance. The second phase consists of safety and environmental factors, in accordance with NRC requirements and guidance. The second phase consists of safety,
environmental, safeguards and antitrust reviews undertaken by the
NRC staff. Third, a safety review is conducted by the independent
Advisory Committee on Reactor Safeguards (ACRS); this review isrequired by law. Fourth, a mandatory public hearing is conductedby a three-member Atomic Safety and Licensing Board (ASLB),
which then makes an initial decision as to whether the permit should
be granted. This decision is subject to appeal to an Atomic Safety andLicensing Appeal Board (AS LAB) and could ultimately go to theCommissioners for final NRC decision. The law provides for appeal
beyond the Commission in the Federal courts.
As soon as an initial application is accepted, or "docketed," by the
NRC, a notice of that fact is published in the Federal Register, andcopies of the application are furnished to appropriate State and local
authorities and to a local public document room (LPDR) establishedin the vicinity of the proposed site, as well as to the NRC-PDR in
Washington, D.C. At the same time, a notice of a public hearing ispublished in the Federal Register and local newspapers) which pro.
vides 30 days for members of the public to petition to intervene in theproceeding. Such petitions are entertained and adjudicated by the
ASLB appointed to the case, with rights of appeal by the petitioner tothe ASLAB.
The NRC staffs safety, safeguards, environmental and antitrust
reviews proceed in parallel. With the guidance of the Standard Format (Regulatory Guide 1.70), the applicant for .a construction permit
lays out the proposed nuclear plant design in a Preliminary Safety
Analysis Report (PSAR). If and when this report has been made sufficiently complete to warrant review, the application is docketed and
NRC staff evaluations begin. Even prior to submission of the report,
NRC staff conducts a substantive review and inspection of the applicant's quality assurance program covering design and procurement.
The safety review is performed by NRC staff in accordance with theStandard Review Plan for Light-Water-Cooled Reactors, initiallypublished in September 1975 and updated periodically. This planstates the acceptance criteria used in evaluating the various systems,
components and structures important to safety and in assessing the
proposed site, and it describes the procedures used in performing thesafety review.
The NRC staff examines the applicant's PSAR to determine
whether the plant design is safe and consistent with NRC rules andregulations; whether valid methods of calculation were employed
and accurately carried out; whether the applicant has conducted hisanalysis and evaluation in sufficient depth and breadth to support
staff approval with respect to safety. When the staff is satisfied thatthe acceptance crheria of the Standard Review Plan have been met
by the applicant's preliminary report, a Safety Evaluation Report is
prepared by the staff summarizing the results of their review regarding the antiCipated effects of the proposed facility on the publichealth and safety.
Following publication of the staff Safety Evaluation Report, the
ACRS completes its review and meets with staff and applicant. The
ACRS then prepares a letter report to the Chairman of the NRC

presenting the results of its independent evaluation and recommending whether or not a construction permit should be issued. The staff
issues a supplement to the Safety Evaluation Report incorporating
any changes or actions adopted as a result of ACRS recommendations. A public hearing can then be held, generally in a community
near the proposed site, on safety aspects of the licenSing decision.
In appropriate cases, NRC may grant a Limited Work Authorization to an applicant in advance of the final decision on the construction permit in order to allow certain work to begin at the site, saving
as much as seven months time. The authorization will not be given,however, until NRC staff has completed environmental impact andsite suitability reviews and the appointed ASLB has conducted apublic hearing on environmental impact and site suitability with afavorable finding. To realize the desired saving of time, the applicant
must submit the environmental portion of the application early.
The environmental review begins with a review of the applicant's
Environmental Report (ER) for acceptability. Assuming the ER is
sufficiently complete to warrant review, it is docketed and ananalysis of the consequences to the environment of the construction
and operation of the proposed facility at the proposed site is begun.
Upon completion of this analysis, a Draft Environmental Statement
is published and distributed with specific requests for review andcomment by Federal, State and local agencies, other interested parties and members of the public. All of their comments are then taken
into account in the preparation of a Final Environmental Statement.
Both the draft and the final statements are made available to the
public at the time of respective publication. During this same time
period NRC is conducting an analysis and preparing a report on site
suitability aspects of the proposed licensing action. Upon completionof these activities, a public hearing, with the appointed ASLB
presiding, may be conducted on environmental and site SUitability
aspects of the propsed licensing action (or a single hearing on both
safety and environmental matters may be held, if that is indicated).
The antitrust reviews of license applications are carried out by the
NRC and the Attorney General in advance of, or concurrently with,other licensing reviews. If an antitrust hearing is required, it is held
separately from those on safety and environmental aspects.
. About two or three years before construction of .the plant isscheduled to be complete, the applicant files an application for an
operating license. A process similar to that for the construction permit is followed. The application is filed, NRC staff and the ACRS
review it, a Safety Evaluation Report and an updated Environmental
Statement are issued. A public hearing is not mandatory at this stage,
but one may be held if requested by affected members of the public
or at the initiative of the Commission. Each license for operation of a
nuclear reactor contains technical specifications which set forth theparticular safety and environmental protection measures to be imposed upon the facility and the conditions that must be met for thefacility to operate.
Once licensed. a nuclear facility remains under NRC surveillance
and undergoes periodiC inspections throughout its operating life. In
cases where the NRC finds that substantial, additional protection isnecessary for the public health and safety or the common defense andsecurity, the NRC may require "backfitting" of a licensed plant, thatis, the addition, elimination or modification of structures, systems Or
components of the plant.

==================================================65related to the accident at Three Miie Island, substantial delays were encountered in the review of applications for operating licenses for nuclear power plants.
To help alleviate this situation, the NRC sought thehelp of technical experts in other Government agencies
on a temporary basis, under interagency contracting
arrangements.
During the latter part of 1979, the Department ofEnergy made available technical specialists fromseveral national laboratories to assist in technicalreviews of applications. Review teams have been
established at the Argonne National Laboratory, theIdaho National Engineering Laboratory, the Savannah River Laboratory, the Oak Ridge National
Laboratory, the Los Alamos Scientific Laboratory, thePacific Northwest Laboratory, the Battelle C;olumbus
Laboratory, the Lawrence Livermore Laboratory,
and the Energy Technology Engineering Center.
Technical assistance has also been obtained from the
U.S. Geological Survey, the U.S. Army Corps ofEngineers, and the Naval Research Laboratory.
These resources will be used during fiscal year 1980
to supplement NRC staff resources for the review oflicense applications, pending recruitment and hiringof additional personnel authorized by Congress.

Licensing Reactor Operators
The safety of a nuclear facility depends not only on
its design but on the qualifications of the people who
operate it. To assure that the people in charge of eachnuclear power plant are capable of directing and performing the activities necessary to reactor operation,
the NRC requires each individual who handles thecontrols of the reactor to be licensed. The requirements for issuance of operators' licenses are set
forth in 10 CFR Part 55. Two types of licenses areissued by the NRC: one for "operators" ,and one for"senior operators." During fiscal year 1979, the NRC
issued 212 new operator licenses, 256 renewals, and 26
amendments, bringing the number of operator licenses
in effect on September 30, 1979 to 992. During thesame period, 184 new licenses, 434 renewals, and 36
amendments were issued for senior operators, bringing
the total to 1,437 in effect.
TMI Related Activities. Following the accident atThree Mile Island Unit 2 (TMI), members of the
Operator Licensing Branch were assigned to theLessons Learned Task Force, Bulletins and Orders
Task Force, and TMI Support Task Force to determine
the role of nuclear power plant operators in the TMI-2
accident, to assist in the development of recommendations for the upgrading of operator training requirements, and to review and recommend changes tonormal, abnormal" and emergency procedures. A
number of these proposals have been submitted to theCommission for review. The activities of these task
forces and other groups dealing with the causes andconsequences of the TMI accident, both within andoutside of the NRC, are covered at length in thepreceding chapter of this report.

UNRESOL VED SAFETY ISSUES
Section 210 of the Energy Reorganization Act of1974, as amended, reads as follows:"Unresolved Safety Issues Plan""Section 210. The Commission shall develop aplan for providing for specification and analysis ofunresolved safety issues relating to nuclear reactors and shall take such action as may be necessary
to implement corrective measures with respect tosuch issues. Such plan shall be submitted to theCongress on or before January 1, 1978 and progress reports shall be included in the annual
report to the Commission thereafter."
In response to this reporting requirement, the NRC
provided a report to the Congress, NUREG-0410, inJanuary of 1978 describing the generic issues programof the Office of Nuclear Reactor Regulation (NRR)
that had been implemented early in 1977. The NRR
program described in NUREG.0410 provides for theidentification of generic issues, the assignment ofTable 1. Nuclear Power Plant Licensing Actions-Fiscal Year 1979
CONSTRUCTION PERMITS
Facility

Date Issued

Location

Tennessee Valley Authority

Yellow Creek 1 & 2

11-29-78

Tishomingo County, Miss.

Long Island Lighting Co.

Jamesport 1 & 2

1-4-79Suffolk County, N.Y.

Applicant(No Limited Work Authorizations or Operating Licenses for nuclear power plants were issued during FY 1979.)

66===================================================
priorities, the development of detailed task actionplans to resolve the issues, the projections of dollar andman~power costs, continuing high level management
oversight of task progress, and public dissemination of
information related to the tasks as they progress.
The 1978 NRC Annual Report described the NRC's
progress towards resolving those issues addressed in the
NRR program that had been identified as "Unresolved
Safety Issues" (p. 19). Seventeen "Unresolved Safety
Issues" were identified, to be addressed by 22 generictasks. Three of these generic tasks have now beenreported as complete.

can be attributed in large measure to the temporary
staff reassignments occasioned by the TMI-2 accident.
The technical scope of several of the tasks also required
some further definition as a result of the accident. The
impacts of the accident on those Unresolved Safety
Issue tasks which were affected are discussed briefly in
the individual task progress reports below.
In an effort to minimize the impact of the TMI-2 accident on these tasks, the Office of Nuclear Reactor
Regulation established, in June of 1979, an interim
organization specifically assigned to continue work on"Unresolved Safety Issues." This interim structure wasstill in place as of January 1, 1980.

Evaluation Process

Identification of New Issues

The definition of an "Unresolved Safety Issue"
developed by the NRC for use in identifying issues that
require reporting to the Congress (pursuant to Section210) is as follows:"An Unresolved Safety Issue is a matter affecting
a number of nuclear power plants that poses important questions concerning the adequacy of existing safety requirements for which a final resolution has not yet been developed and that involves
conditions not likely to be acceptable over the
lifetime of the plants affected."
The process used to determine which issues met thedefinition of an "Unresolved Safety Issue" was described briefly in the 1978 NRC Annual Report and in considerably more detail in NUREG-0510, "Identificationof Unresolved Safety Issues Relating to Nuclear Power
Plants-A Report to Congress." (NUREG-0510 accompanied the 1978 Annual Report when it wastransmitted to the Congress in Jan,-,ary of 1979.) The
review process included a systematic review of over
130 generic issues. As an aid to this review, an evaluation was made of the subject areas involved according
to their relative importance from the standpoint ofpublic risk. This risk-based characterization was used
together with a substantial body of additional information (e. g., heavy weight was given to issues arisingfrom events reported to the Congress as "Abnormal
Occurrences") to determine which issues qualified forreporting to Congress.
Table 2 provides a listing of each of the "Unresolved
Safety Issues" and related generic tasks identified in
the 1978 NRC Annual Report. It also provides last
year's projected dates for issuing NRC staff reports andthe corresponding dates as of December 1, 1979.
As indicated in Table 2, three reports providing thestaff's resolution of three "Unresolved Safety Issues"
were issued for public comment as of January 1, 1980.
Four more staff reports addressing four more
"Unresolved Safety Issues" are expected to be issued by
the end of February 1980. A number of the tasks haveundergone schedule slips in 1979. These schedule slips

Although a number of safety-related issues came tolight in 1979 as a result of the TMI-2 accident andother events, the NRC staff has not been able to perform an in-depth review to identify and evaluate newissues. Therefore, no new "Unresolved Safety Issues"
have been defined for reporting in 1979. As of January
1, 1980, NRC efforts were being concentrated on implementing new TMI-related requirements on
operating plants and on identifying, defining andscoping additional TMI-related issues and tasks.
Several broad program areas where issues and tasks
are being scoped will likely result in designation of
new Unresolved Safety Issues. These program areas include the following:(1) Man-machine interface and control-roomdesign.
(2) Qualification and training of operation,
maintenance and supervisory personnel.
(3) Off-site emergency response, emergency planing, and action guidelines.
(4) Siting policy, including compensatory designand operating provisions for plants in areas
where evacuation would be difficult.
(5) Systems reliability and interactions.
(6) Consideration in licensing requirements of accidents involving degraded or melted fuel.
The NRC staff performed a cursory review of anumber of candidate issues from sources other than
TMI accident investigations, including a review of
events reported as Abnormal Occurrences in 1979.
None of these issues was judged to be of such safety
urgency and importance as to require reporting in advance of the staff's and the Commission's in-depth andsystematic review of all candidate issues. Such asystematic and in-depth review will be performed in
1980, after the major recommendations of the major
TMI investigations are available. A special report will
be provided to the Congress by July of 1980, describing the review and the new issues to be designatedUnresolved Safety Issues.

==================================================67

Table 2.

Schedules for Tasks Addressing Unresolved Safety Issues
Schedule jor
Issuing Staff
Report as ofJanuary 1,

No.

Unresolved Safety Issue

Schedule for Issui~
Staff Report in 19 8
NRC Annual Report

A-I
A-2
A-3
A-4

Water Hammer
Asymmetric Blowdown Loads
PWR Steam Generator Tube Integrity

1980
Early 1979
Early 1980

August 1981
January 1980
May 1980.

BWR Mark I and Mark II Pressure Suppression Containments

A-7-0ctober 1979
A-8-0ctober 1980
A-39-0ctober 1979
Early 1979
Late 1979
July 1979
August 1979

February 1980
November 1980
March 1980
April 1980
February 1980
December 1980
IssuedNovember 1979
April 1980
May 1981
IssuedDecember 1979
January 1980
February 1980
March 1981
IssuedOctober 1979
Not Scheduled·
Not Scheduled·

Task

A-5
A-7
A-8

A-39
A-9
A-lO
A-ll
A-12

Anticipated Transients Without Scram
BWR Nozzle Cracking
Reactor Vessel Materials Toughness
Steam Generator and Reactor Coolant Pump Supports

A-17

Systems Interactions

A-24

Qualification of Class IE Safety-Related Equipment

A-36
A-40

Control of Heavy Loads Near Spent Fuel
Seismic Design Criteria

A-42

Pipe Cracks in Boiling Water Reactors

Early 1979
Phase 1-1979
Phase 11-1981
Not Scheduled

A-43
A-44

Containment Emergency Sump
Station Blackout

Not Scheduled
Not Scheduled

Phase I-September 1979
Phase II-September 1980
1979

1980

·Task Action Plan unqer development when TMI accident occurred; it is anticipated that the task can be completed in 1982.

Progress Reports~

Background information and progress reports foreach of the Unresolved Safety Issues listed in Table 2
are provided below. Progress reports on the staffs efforts on tasks reported as unresolved in last year's annual report are also provided. As indicated above,
NRC staff reports have been issued for three
Unresolved Safety Issue tasks as of January 1, 1980.
These tasks are: Task A-I2, "Fracture Toughness of
Stearn Generator and Reactor Coolant Pump Supports"; A-24, "Qualification of Class IE SafetyRelated Equipment"; and Task A-42, "Pipe Cracks in
BOiling Water Reactors." The reports describe thetechnical studies conducted by the NRC staff or its
contractors and the safety conclusions that provide the

NRC staffs resolution of each of these safety issues.
Broad public and industry comment is being solicitedon these three reports.

Water Hammer
Water hammer events are intense pressure pulses influid systems (such as commonly experienced when
rapidly closing a water faucet) and they often occur innuclear power plant fluid systems. In the past few
years, over 200 incidents involving water hammer innuclear power reactors have been reported. These incidents have involved many types of fluid systems, including stearn generator feed-rings, feedwater andstearn supply piping, residual heat removal systems,
emergency core cooling systems, containment spray

68

systems, and service water systems. Water hammer
can have various causes, such as the rapid condensation of steam pockets, steam-driven slugs of water,pump startup with partially empty lines, and rapid
valve motions. Most of the damage has been relatively
minor, though there have been several cases of failure
or partial failure of system piping.
While no water hammer incident has resulted in therelease of radioactivity outside of a plant, the concern
is that water hammer could result in the failure of apipe in the reactor coolant system or disable a system
required to cool the plant after a reactor shutdown.
The means to prevent one particular type of water
hammer caused by the rapid condensation of steam in
the steam generator feed- rings of some pressurizedwater reactors are being instituted. In addition, applicants with new steam generator designs are being
required to demonstrate through test or analysis thatwater hammer will not occur in these designs. Plants
with steam generators-of the top feeding type thatare subject to water hammer-are being required tomodify the feed-rings and/or test the systems to assure
water hammer will not occur. Other actions to correct
the specific causes of water hammer in other nuclear
power plant systems are also being required.
Under Generic Task A-I, the potential for water
hammer in various systems is being evaluated and appropriate requirements and systematic review procedures are being developed to ensure that water hammer is given appropriate consideration in all areas oflicensing reviews. A technical report, NUREG-0582,"Water Hammer In Nuclear Power Plants," providing
the Iesults of an NRC staff review of water hammer
events in nuclear power plants and stating current
staff licensing positions, was published in July 1979.
Issuance of this report completes a major subtask ofGeneric Task A-I.
In addition, seven technical reports on water hammer have been issued by NRC contractors participating in Task A-I. Issuance of these reports completes
four other major subtasks of Generic Task A-I. Collectively, these seven reports have provided: (1) a review
and evaluation of actual and potential water hammer
events in nuclear power plants, (2) analytical methods
and calculational procedures to be used in the evaluation of water hammer incidents, and (3) current stateof-the-art information on water hammer. At the closeof the report period, a report summarizing these NRC
sponsored water hammer studies had been completed
in draft form for review. Issuance of this report willcomplete another major subtask.
The need for additional work to evaluate the safetysignificance of various water hammer scenarios has
been identified, and the scope of Task A-I has been expanded to include such studies. This need, combinedwith the manpower impacts of the Three Mile Island
Unit 2 accident, resulted in a schedule slip of about 7

months in the projected completion date for Task A-I
to August 1981.

Asymetric Blowdown Loads
On the Reactor Coolant System
In the very unlikely event of a rupture of the
primary coolant piping in light water reactors, large
non-uniformly distributed loads would be imposed
upon the reactor vessel, reactor vessel internals, andother components in the reactor coolant system. The
potential for such asymmetric loads, which result fromthe rapid depressurization of the reactor coolant
systems, was first identified in 1975 and thus was nut
considered in the original design of some facilities.
Details on the safety significance of this issue and actions taken by NRC and industry prior to fiscal year
1979 in response to it may be found in the 1978 NRC
Annual Report, pp. 21-24.
Plant modifications to assure that the postulatedloads are accommodated have been implemented latein the construction stage of several plants and havebeen proposed and are under staff review for some
operating plants. For plants still under operating
license review, the NRC staff requires that plantspecific analyses and any necessary plant modifications be completed prior to issuance of an operating
license. The staff reviewed and approved topicalreports from the vendors of pressurized water reactors(PWRs) explaining their generic approaches to thecalculation of the asymmetric loads in a loss-of-coolant
accident (LOCA). Lead plant evaluations and sensitivity studies were also conducted by the staff. A
report providing the NRC staffs resolution of this issue
is expected to be issued by the end of January 1980.
This report will provide acceptance criteria andguidelines for use in plant-specific analyses.
Plant-by-plant implementation of the results of
Task A-2 actually began with a request for plantspecific analyses in January 1978. All licensees with
operating PWRs are in the process of evaluating theirplant's capacity to sustain asymmetric LOCA loads.
The PWR owners' groups have met with the staff
periodically to present interim results and progress. Allof the requested evaluations are scheduled to be completed by early 1980 at which time NRC staff review
and implementation of plant modifications will begin.
Asymmetric blowdown loads may also be importantin boiling water reactor (BWR) plants, but they are expeeted to have lesser safety significance than in PWRs,
because of the lower operating pressures in BWRs. A
plan for resolving the matter for BWR plants will be
developed by the NRC staff and pursued separatelyfrom the PWR issue.

69

PWR STEAM GENERATOR

Slj~:tt---- STEAM OUTLET TO

TURBINE GENERATOR

The buildup of corrosion deposits between thesteam generator tubes and the tube support
plates, in addition to constricting the tubes, exerts stresses on the tube SUPIl0rt plates. The
stresses cause hourglassing of the normally rectangular internal bypass flow holes locatedbetween the innermost tube rows.

TUBE BUNDLE

TUBE SUPPORTPLATE//..._ _-....:r=-~

PRIMARY COOLANT
OUTLET
PRIMARY COOLANT
INLET

PWR Steam Generator Tube Integrity
The heat produced in the reactor at a nuclear power
plant is used to convert water into the steam which
drives the turbine-generators. In plants employing
pressurized water reactors, the primary coolant water,which extracts heat by circulating through the reactor
core and is radioactive, is kept under pressure sufficient to prevent boiling. This high-pressure water
passes through tubes around which a secondary coolant (also water, but not radioactive) is circulating
under somewhat lower pressure. The water in thesecondary system boils and produces steam to drive theturbine generators. The assembly in which the heattransfer takes place is the steam generator. The tubes
within it are an integral part of the primary coolant
boundary, keeping the radioactive primary coolant ina closed system and isolated from the environment.
The primary concern is the maintenance of steam
generator tube integrity during both normal operationand postulated accident conditions. Another concern
is that the increased steam generator tube inspections
and repairs have resulted in significant increases in occupational exposures to workers.
A discussion of the specific problems associated withsteam generator tube integrity that were occurring at

operating reactors was provided in the 1977 and 1978
NRC Annual Reports on pp. 95 and 22, respectively. A
more detailed discussion of steam generator operating
experience is provided in NUREG-0523, "Operating
Experience with Recirculation Steam Generators,"
published in January 1979 and in NUREG-0571,"Operating Experience with Once Through SteamGenerators," to be published in fiscal year 1980.
The significant developments in Westinghouse andCombustion Engineering steam generators, sinceAugust 1978, were the following:• Steam generator replacement at Surry Unit 2 isessentially completed. Replacement is planned for
Surry Unit 2, Turkey Point Units 3 and 4, andPalisades. In the interim, the units are operating
under restrictions imposed by the NRC.
• Condenser retubing to reduce in-leakage ofseawater and the installation of full-flow
demineralizers in the secondary coolant system,
to remove any chlorides which might leak, hasretarded the rate of tube denting at Millstone
Unit 2.
• Yankee Rowe performed a 100 percent inspectionof Steam Generators 1 and 4 in November 1978.

70

Fifteen and 12 defective tubes were discovered in
Steam Generators 1 and 4, respectively. The
mode of degradation was secondary side wastage.
A leak rate of approximately 125 gallons-per-day
existed prior to the inspection. The Unit, which
has stainless steel tubing, converted from
phosphate to AVT secondary water treatment in1968.
• Point Beach Unit 1 was required to shut down on
September 20, 1978 and March 12 and August 5,
1979 because of steam generator tube leaks. The
cause of the leaks was cracking of the tubes in thecrevice between the tubes and the tubesheet. The
cracking was a result of caustic stress corrosion. A
100 percent inspection of both steam generators
was performed. The inspection revealed 52 defective tubes in Steam Generator A and 45 defective
tubes in Steam Generator B. All the cracks werelocated within the tubesheet and are therefore notconsidered a significant safety concern.
Oconee Units 1, 2 and 3 and Crystal River Unit 3 are
the only Babcock and Wilcox units which have had
steam generator tube leaks. Tubes in one localized
area of the Oconee Unit 1, 2 and 3 steam generators
have failed because of cracks of unknown origin propagated circumferentially by flow-induced vibration.
The status in the B&W steam generators, sinceAugust 1978, is the following:• The Oconee Units have not had a steam generator
leak related to a fatigue crack since April 1978.
However, a steam generator tube leak (not believed to be related to a fatigue crack) occurred on July 24, 1979.
• Crystal River Unit 3 was shut down for a steam
generator tube leak on August 19, 1979. The leakis believed to be through a steam generator tubeto-tubesheet weld which was damaged when aburnable poison rod assembly broke up in Marchof 1978.
• A demonstration tube sleeving program was initiated by Duke Power Company at the Oconee
Units. The tube sleeves will be installed to change
the viorational characteristics of the tubes anddecrease the dynamic stress and the susceptibilityof the tubes to fatigue cracking. They will notserve as part of the primary coolant boundary.
• An additional degradation mechanism, defined as~n "erosion-corrosion" phenomenon and resulting
III tube wall thinning, has been identified at
Oconee and other B&W units.
Plant technical specifications require routine inservice inspection of steam generators to be performed
every 12 to 24 months. The NRC has imposed license
conditions on plants with severely degraded steam
generators to increase the required frequency of in-spection. The conditions also require that following inspection of steam generators and completion of any
necessary repair programs by the licensees, the NRC
must approve or concur in the restart of each severely
affected facility. To date, the units severely affectedby tube wall thinning have completed inspection andrepair programs and received NRC approval for
operation for limited time periods. Safe operation isassured by the imposition of strict conditions requiring
the plugging of affected tubes and restricting
allowable leak rates during operation. While the NRC
continues to closely monitor and evaluate the acceptability of continued operation of plants experiencing
steam generator tube problems, it is proceeding withthree generic tasks in the NRC program for the resolution of generic issues (specifically, Generic Tasks A-3,
A.4, and A-5, addressed to the problems ofWestinghouse, Combustion Engineering, and Babcock
and Wilcox steam generators, respectively.)
The Task Action Plans for these tasks have beencombined in a single plan encompassing all threetasks. The approach taken in the Task Action Plan is tointegrate technical studies in the three areas of systemanalyses, inservice inspection, and tube integrity inorder to establish improved criteria by which to ensure
safe and reliable steam generator operation. The purpose of the system analyses is to evaluate the conse'"
quences of failures involving different numbers ofsteam generator tubes during postulated accident conditions (LOCA and main-stearn-line break, or
MSLB) -considering predicted fuel behavior,
emergency core cooling system (ECCS) performance,
radiological consequences, and containment response.
The results will be used to define the tolerable level ofsteam generator tube leakage during postulated accidents. The major emphasis in the inservice inspection portion of the tasks is to develop a statistically based inservice inspection program that will provide
assurance that no more than the tolerable level of tube
leakage, defined by the system analyses, would occur
in an accident. The tube integrity portion of the tasks
is primarily concerned with experimental verificationof the tube behavior during postulated accidents,
development of tube plugging criteria, and definitionof operating procedures for minimizing tube degradation.
The statistical analyses of inservice inspection programs, which is being performed parametrically, wasscheduled to be complete in February 1980. The
system analyses and tube integrity evaluation werescheduled for completion in early 1980. The results of
the Task Action Plan will be: (1) tube plugging criteria
based on new experimental data, (2) statistically based
inservice inspection methods, (3) recommendations for
improved methods of operation, and (4) recommendations for design improvements for new plants. These
will be described in an NRC staff report scheduled to
be issued for comment in May 1980.

=================================================71
BWR Mark I and Mark II
Pressure Suppression Containments
Boiling water reactor pressure suppression containments designed by the General Electric Company
utilize a large mass of water as the principal heat sinkby which to condense the steam and absorb the energythat might be released from the reactor in the unlikely
event that certain pipes in the primary system should
fail. As postulated, the absorption of energy by thestored water reduces the potential buildup of pressure
inside the containment~ and that in turn reduces thedriving force that might lead to a release to the environment of fission products that may have beenreleased from the reactor cQre.
In the course of performing large scale testing of anadvance design pressure-suppression containment(Mark III), and during in-plant testing of Mark I containments, new suppression pool hydrodynamic loads
were identified which had not been explicitly includedin the original Mark I or Mark II containment designbasis. These additional loads result from the dynamic
effects of drywell air and steam being rapidly forced
into the suppression pool during a postulated LOCA,
and from the suppression pool response to various
modes of safety relief valve (SRV) operation generally
associated with operating conditions during an accident. Since these new hydrodynamic loads had notbeen explicitly considered in the original design of theMark I and Mark II containments, the NRC staff
determined that a detailed reevaluation of these containment system designs was required. The affectedutilities formed Mark I and Mark II Owners' Groups
and drew up both short-term and long-term programsfor resolution of the pool dynamic problems for theirrespective containment designs. The programs include
a n~m ber of comprehensive experimental andanalytical programs to establish pool dynamic loads,
load combinations, and design criteria.
The NRC staff has identified and initiated a numberof generic tasks to review and evaluate the results of
the industry programs and to develop criteria forlicensing actions on individual plants using the Mark I
and Mark II containment designs. These generic tasks
are included in the NRC Program for Resolution ofGeneric Issues. Specifically, they are Task A-6, "MarkI Short-Term Program"; Task A-7, "Mark I LongTerm Program"; Task A-8, "Mark II Containment
Program"; and Task A-39, "Determination of Safety
Relief Valve Pool Dynamic Loads and Temperature
Limits for BWR Containment."
The objectives of the Mark I Short-Term Program
Were: (1) to examine the containment system of each
BWR facility with a Mark I containment design toverify that it would maintain its integrity and functional capability when subjected to the most probablehydrodynamic loads induced by a postulated designbasis loss-of-coolant accident; and (2) to verify thatlicensed Mark I BWR facilities may continue tooperate safely, without undue risk to the health andsafety of the public, while a methodical, comprehensive long-term program is conducted.
The NRC determined that, for the Short-Term Program, "maintenance of containment integrity andfunction" would be adequately assured if a safety factor to failure of at least two were demonstrated to exist
for the weakest structural or mechanical component in
the Mark I containment system (that is, if thecalculated stresses in all components of the affectedcontainment structure were shown to be less than onehalf the stress which would cause the component tolose its structural integrity). The NRC concluded thatthe objectives of the Short-Term Program had beensatisfied and documented the basis for this conclusionin the "Mark I Containment Short-Term Program
Safety Evaluation Report," NUREG-0408, dated
December 1977. (Thus Task A-6 was completed in
December 1977.)
The objectives of the Mark I Long-Term Program
are: (1) to establish design b~is (conservative) loads
that are appropriate for the anticipated life (40 years)of each Mark I BWR facility, and (2) to makewhatever plant changes may be required to restore theoriginal intended design safety margins for each MarkI containment system. The industry program includes
experiments and calculations designed to provide a
detailed basis for hydrodynamic load definition andstructural assessments. The generic aspects of the program are described in a Plant Unique Analysis Applications Guide (General Electric Topical Report
NEDO-24583), which was submitted by the Mark I
Owners' Group in February 1979, and a Load Definition Report (General Electric Topical Report
NEDO-21888), which was submitted, in two parts, in
December 1978 and March 1979. These reports
described the proposed load definition and assessment
techniques for Mark I containments. They werereviewed by the NRC staff, who issued a set of acceptance criteria for the generic assessment techniques in
September 1979. A Safety Evaluation Report describing the stafr s review and the bases for the acceptancecriteria was scheduled to be issued in February 1980,
marking the completion of Generic Task A-7.
Subsequently, each utility will be required to perform a plant-unique analysis using approved load
definition and structural analysis techniques todemonstrate conformance with the structural acceptance criteria. The scheduled completion date for theMark I Long-Term Program-including the issuanceof license amendments and the implementation of any
plant modifications necessary to satisfy the Long-Term
Program structural acceptance criteria-is December
1980. To maintain this schedule, a number of utilities

72

BWR PRESSURE SUPPRESSION CONTAINMENTS

MARK IMARK IIMARK IIIThe design objective of BWR containment systems is to condense
the steam released during certain, postulated accidents to limit therelease of fission productS within the reactor building and to provide
water for emergency cooling systems. NRC's Generic Task A-39
entails the review and evaluation of an industry-based program toestablish pool dynamic loads for BWR Mark I, II and III pressure
suppression containment designs.

have undertaken plant modifications prior to the completion of their plant-unique analysis. This action has
been considered necessary to minimize the potential
for unduly long extensions of exemptions or extended
plant outages later in the program. Similarly,
modifications to components external to the containment (e. g., support structures) have and are being
conducted during normal plant operation.
The Mark II Owners' Group developed a program
consisting of a number of analytical and experimental
tasks to support these pool dynamic loads application
methods. They divided the overall program into twoparts: A Lead Plant Program and a Long-Term Program. The objective of the Mark II Lead Plant Program was to establish design basis (conservative) loads
appropriate for the anticipated life of each Mark II
BWR facility. The Mark II owners' specification of theLead Plant Program loads are described in Revision 2of the Dynamic Forcing Function Report, NEDO21061-P, and in several application memoranda. The
tasks comprising the Lead Plant Program and LongTerm Program are listed in NUREG-0487.
The staff has reviewed and evaluated the pool
dynamic loads associated with a postulated large lossof-coolant accident proposed by the Mark II Owners'
Group to determine their acceptability for use inplant-unique analyses for the lead plants. The Mark II
Lead Plant Program was essentially completed in October 1978 with the publication of NUREG-0487,"Mark II Containment Lead Plant Program Load
Evaluation and Acceptance Criteria." The lead MarkII plants proposed several exceptions to these criteria.
The staff has reviewed these proposed exceptions andhas found most of them acceptable. A safety evaluation of these exceptions and their bases was to be
discussed by the staff in a supplement to
NUREG-0487.
The purpose of the Mark II Long Term Program isto refine the load definitions used in the Lead Plant
Program and to support by additional experimental
and theoretical work, a reduction in some of theseloads for use in the evaluation of the Mark II plants
following the lead plants. The Mark II owners'
technical program includes a total of 101 tasks. This
program is about 70 percent complete. During the past
year, the Mark II owners have completed a number ofthe Long-Term Program tasks and numerous technicalreports have been provided for NRC review. Late in
1979, the Mark II owners modified the Long Term
Program to include several new analytical and test
programs. The addition of these new tasks could delay
the Mark II Long Term Program review beyond the
projected completion date of January 1981.
Under Generic Task A-39, the NRC staff will review
and evaluate the results of industry experimental andanalytical programs to establish and justify the safety
relief valve-related pool dynamic loads for BWR Mark73

I, II, and III containment designs. The results ofGeneric Task A-39 will be an integral part of the final
acceptability of these designs. The portions of thisgeneric task related to the Mark I and Mark II containments are currently scheduled to be completed in
March 1980.

Anticipated Transients Without Scram
Nuclear plants have safety and control systems tolimit the consequences of abnormal operating condiw
tions transients. Some deviations from .normal
operating conditions may be minor; others, occurring
less frequently, may impose significant demands onplant equipment. "Anticipated Operational Occurrences" or "Anticipated Transients" are defined (10CFR Part 50, Appendix A) as "those conditions of normal operation which are expected to occur one ormore times during the life of the nuclear power unit
and include but are not limited to loss of power to allrecirculation pumps, tripping of the turbine generator
set, isolation of the main condenser, and loss of all offsite power." In some anticipated transients, rapidly
shutting down the nuclear reaction (initiating a"scram"), and thus rapidly reducing the generation ofheat in the reactor core, is an important safetymeasure. If there were a potentially severe anticipatedtransient and the reactor shutdown system did notfunction as designed, then an "anticipated-transientwithout-scram," or ATWS, would have occurred.
This issue has been discussed throughout the NRC
and AEC and the nuclear industry for a number ofyears. Details on the safety significance of the issue
and actions taken by NRC and industry prior to fiscal
year 1979 in response to it may be found in the 1978
NRC Annual Report, pp. 27 and 28.
On the basis of discussions with senior NRC
management, the Advisory Committee on Reactor
Safeguards, and industry representatives, and thereview of the Lewis Committee report on the Reactor
Safety Study, the NRC staff in December 1978 proposed a combination of preventative and mitigativemeans of prOViding protection from ATWS events. In
this supplement, the NRC staff proposed different
types of plant modifications. The design alternatives
which were proposed take into consideration the statusof the plants-whether operating, under construction
or nearly ready for operation-and questions of practicability, including the cost of such modifications.
In order to confirm that these alternatives provided
the needed level of safety, the industry was required toprovide the necessary confirmation analyses and thestaff originally intended to make its recommendations
to the Commission in the spring of 1979. The Three
Mile Island Unit 2 accident affected these plans inseveral ways. First, both industrY and NRC staff manpower were diverted from ATWS work; second, theThree Mile Island event scenario indicated that a
number of aspects of the ATWS accident evaluation
required reconsideration, especially for PWRs.
The shortage of available industry manpower
delayed several of the required submittals of confirmation analyses. The result was a substantial slip in the
projected completion date for the ATWS task. NRC
staff manpower was partially restored in June 1979and meetings were held with industry representatives
in July and August of 1979 to discuss the impacts of theThree Mile Island Unit 2 accident on their respective
ATWS evaluations. As of January 1, 1980, the NRC
staff planned to propose an ATWS rule to the Commission by April 15, 1980, with a goal of issuing a final
ATWS rule by December 1980.

BWR Nozzle Cracking
Over the last several years, inspections at 22 of the
31 boiling water reactor (BWR) plants licensed for
operation in the U.S. have disclosed some degree ofcracking in the feedwater nozzles of the reactor vessel
at 18 facilities. One facility has not yet accumulatedSignificant operating time and has, therefore, not yet
been inspected.
The feedwater nozzles are an integral part of the
primary pressure boundary of the reactor coolant.system and the second barrier (after the fuel cladding)to the release of radioactive fission products. All of therepaired BWR feedwater nozzles met the ASME
pressure vessel code limits, however, and no immediate action was necessary. Because only relatively
small amounts of metal have been removed by repair
operations, there has been no significant reduction insafety margins. Nevertheless, the cracking is potentially serious because:• Excessive crack growth could lead to impairmentof pressure vessel safety margins.
• The design safety margin could also be reduced
by excessive removal of nozzle reinforcement
while grinding out cracks, and repair by welding
would be complicated.
• The exposure· to radiation of the personnel
performing inspection and repair tasks can beconsiderable.
• The repair of these kinds of cracks can result in
considerable shutdown time at the plant affected.
The reactor vendor (the General Electric Company)
and the NRC have concluded from their separate
studies that the cracks are initiated by rapid fluctuations in water temperature on the inside surface of thenozzles during periods of low feedwater temperaturewhen flow may also be unsteady and perhaps intermittent. The cracks then grow deeper as a result of
operational startup and shutdown· cycles or otheroperationally induced transients. The stainless steel
cladding exhibited less resistance to crack initiation
than the underlying low-allow steel.

74

FEEDWATER NOZZLENOZZLE BLENDRADIUS CRACKS

o~R~,-,-,,:Pr~~;;:===~:......L:::LEAKAGEE IBYPASS"-FEEDWATER-~'"(Coldl~~-~Cracks in nozzles of the feed water and control rod drive lines of
BWR reactor pressure vessel have bt,·tm studied by the vendor(General Electric) and the NRC staff for several years. Evidence

indicates that abrupt and wide fluctuations in water temperatures(see diagram above) are the initial causes of cracking. Photo aboveshows such cracks.

The vendor has performed extensive analysis andtesting to confirm the suspected cause of the cracking
and to develop possible long-term solutions-a newly
designed sleeve, removal of the stainless steel cladding,
reduction of the temperature differential at the nozzle,
or some combination of these. The licensees involved
have increased the number and extent of in-service inspections of feedwater nozzles, with careful repair andreinspection where cracks were found. The vendor advised these licensees to closely monitor startup andshutdown procedures in an effort to substantiallyreduce the time during which cold feedwater is being
injected into the hot pressure vessel.
In a closely related area, the NRC was informed in
March 1977 by the General Electric Company that acrack had been found in the nozzle of the control rod
drive (eRD) return line in a reactor vessel. The CRD
return line nozzles are the openings in BWR pressure
vessels through which the high pressure water in excessof that needed to operate and cool the CRDs is returned to the pressure vessel. The cracks resembled thosefound in the feed water nozzles and seemed to be theresult of the same kind of cyclic thermal stresses thatwere causing feedwater nozzle cracks. The maximum
crack depth has been 0.87 inch.
The NRC staff efforts related to the resolution ofthese two similar issues regarding nozzle cracking inboiling water reactors were consolidated into a single
staff effort, Generic Task A-10, in 1977. Under
Generic Task A-10, the staff issued interim guidance tooperating plants in a report entitled, "Interim
Technical Report on BWR Feedwater and Control
Rod Drive Return Line Nozzle Cracking," in 1977.

The staff has now completed its review of theGeneral Electric studies on feedwater nozzle cracking
and has concluded that the new sparger design-in
conjunction with other remedial measures, such as
clad removal and more appropriate operating procedures-is an effective means of greatly reducing theprobability of crack initiation. The new sparger design
includes flow discharge nozzles, a triple thermal
sleeve, and two piston ring seals in the nozzle bore.
The effectiveness of the new design in minimizing bypass leakage and other problems encountered in theolder designs was confirmed by extensive testing andanalyses by General Electric, including vibration,
thermal-hydraulic, materials, and thermal fatigue
evaluations. Other designs may also be acceptable.
Feedwater system changes, necessary on some low
feedwater temperature plants to assure no cracking
over the lifetime of the plant, are being evaluated on aplant-specific basis. An NRC staff report incorporating guidance for operating reactors and plants underlicensing review is in preparation and is scheduled to
be issued for comment in February 1980.
The resolution of questions regarding the future
selection of improved inservice inspection techniques
and frequency of inspection has been separated fromthe generic task while major industry investigations
continue '(including thermal cracking in a full-size
nozzle mockup to be used in ultrasonic evaluation). A
supplement to the NRC staff report cited above may
be necessary upon completion of these studies. In themeantime, stringent inspection requirements, based
mainly upon dye-penetrating testing, are still in force.
All licensee efforts, such as system and operational

==================================================75
changes, to lengthen the time to crack initiation and toslow crack growth are taken into account in the determination of inspection techniques and criteria.
The CRD nozzle issue will be resolved by a combination of actions which includes nozzle inspection
and repairs and some CRD system notifications. Certain system modifications recommended by General
Electric involved cutting and capping the nozzle andreturn line but that action would reduce the capability
to direct high pressure water through the CRD system
when the vessel is otherwise isolated. Although thissystem is not normally expected to perform this function in safety analyses, the capability played a majorrole in keeping the core covered during the incident at
Browns Ferry Unit 1 on March 22, 1975. As a result of
its review of these modifications, the NRC has concluded that only a limited number of plants will beallowed to modify the CRD system in accordance with
the GE recommendations. Unless the licensees of theremaining plants demonstrate, by testing, that sufficient flow is available to the reactor vessel with thereturn line removed, they will be required to retain
the return line, rerouted to the feedwater line or asimilar suitable connection that doesn't have thepotential for cracking in the reactor vessel nozzle. The
staffs evaluation, conclusions, and guidance on the,CRD return line nozzle issue will also be included in
the February 1980 NRC staff report referred to above.
Plant-specific implementation of the generic licensing positions developed under this task (with the exception of future inservice inspection questions) has
already begun.

tion of pressure and temperature will remain well
below that which might cause brittle fracture of thereactor vessel if a significant flaw were present in thevessel material. The effect of neutron radiation on thefracture toughness of the vessel material is accounted
for in developing and revising these technical specifications over the life of the plant.
For the ,service time and operating conditions
typical of current operating plants, reactor vessel fracture toughness provides adequate margins of safety
against vessel failure. Further, for most plants thevessel material properties are such that adequate fracture toughness can be maintained over the life of theplants. However, results from a reactor vessel
surveillance program indicate that up to 20 older
operating pressurized water reactor pressure vessels
were fabricated with materials that will havemarginal toughness after comparatively short periodsof operation. This issue has been incorporated in theReactor Vessel Material Toughness
Nuclear reactor pressure vessels are required to havean adequate margin of protection against fracture in
the presence of relatively large postulated flaws. Thisrequirement is imposed for the sake of conservatism,
even though extensive, periodiC inservice inspection
programs serve to provide protection against thepresence of such flaws. Fracture mechanics-theengineering method used to establish the failure
margin-employs a quantitative material property
called fracture to'ughness to calculate the conditions
under which catastrophically rapid crack propagation
will occur. Fracture toughness has different values andcharacteristics depending upon the material being
considered. For steels used in nuclear reactor pressure
vessels, three facts are important. First, fracture
toughness increases with increasing temperature. Second, fracture toughness decreases with increasingload rates. Third, fracture toughness decreases withneutron irradiation.
In recognition of these considerations, the technicalspecifications for power reactors set limits on theoperating pressure during heat up and cooldown
operations. These restrictions assure that the combina-

The protective insulation has been pulled aside following thetesting of a weld-rej)air portion of a six-inch thick pressure vessel.
A flaw more than five inches deep and 18 inches long was createdin the area which was then subjected to pressure overloads morethan double the design pressure, without disruptive failure.

76

NRC staffs program for the resolution of generic issues
as Task A-II.
The fundamental goal of Task A-II is to provide an
improved engineering method by which to assess thesafety margin in nuclear reactor pressure vessels and todevelop appropriate criteria for the evaluation of normal, transient, or postulated accident conditions
under the improved method. This method could thenbe used to provide such an assessment for those older
reactor pressure vessels that will eventually havemarginal toughness according to the current method.
Because relatively large amounts of prefracture plastic
deformation can be expected at high temperatureseven in pressure vessel steels of low toughness, the new
evaluation method will employ "elastic-plastic" fracture mechanics concepts. The basis for this improved
methodology is described in NUREG-031I, "A Treatment of the Subject of Tearing Instability," developed
under an NRC-sponsored program at Washington
University. Additional Washington University work
extending the methodology to reactor pressure vessels
was funded by the Department of Energy. The
engineering method developed will account forradiation-induced material degradation.
Task A-II also includes or relies on programs sponsored by the NRC Office of Nuclear Regulatory Research to provide: (1) an improved evaluation ofmaterial degradation mechanisms resulting fromneutron irradiation, and (2) the development of improved testing methods for use in determining theelastic-plastic properties of materials.
Since last year's report, the following has been accomplished:• Although delayed, an elastic-plastic fracture test
method for routine determination of fracture
toughness was developed. Verification of the test
method is underway.
• The elastic-plastic fracture mechanics methods of
NUREG-031I were confirmed by work supported
by an Electric Power Research Institute program,"Methodology for Plastic Fracture."• The methods developed in these programs weresuccessfully used by NRC contractors to analyze
two pressure vessel burst tests reported in the
Heavy Section Steel Technology Program, sponsored by the NRC Office of Nuclear Regulatory
Research.
• The potential for restoring by thermal annealing
the pressure vessel toughness lost by neutronradiation was shown to be impractical.
Significant delays have developed over the past year
as a result of difficulties encountered in extending the
new engineering methodology to reactor pressure
vessels. There is agreement among experts that themethodology can be extended, but it will require asignificantly greater effort than that accomplishedunder the DOE contract referred to above. The staff
will carry out this additional work by contract with
the Oak Ridge National Laboratory. Because of thisproblem, the schedule for completing Task A-II hasslipped about one year, to December 1980.

Fracture Toughness and Potential for
Lamellar Tearing of PWR Steam Generator
And Reactor Coolant Pump Supports
During the course of licensing review for a specificpressurized water reactor (PWR) a number of questions were raised as to (1) the adequacy of the fracture
toughness properties of the material used to fabricate
the reactor coolant pump supports and steam
generator supports, and (2) the potential for failure
due to lamellar tearing of these same supports. The
safety concern is that, although these supports aredesigned for worst-case accident conditions, low fracture toughness or lamellar tearing could cause the support to fail during such accidents. Support failure
could conceivably impair the effectiveness of systems
designed to mitigate the consequences of the accident.
An example of a postulated event sequence of potential
concern would be a large pipe break in the reactor
coolant system which would severely load the supports, followed by a support failure of sufficientmagnitude that a major component such as a steam
generator would be displaced resulting in failure of theemergency core cooling system piping needed to provide cooling water to the core.
Because materials and designs similar to those of the
PWR originally reviewed have been used in otherplants, review of this issue was included in the NRC
Program for Resolution of Generic Issues as GenericTask A-I2.
A consultant was engaged to reassess the fracture
toughness of the steam generator and reactor coolant
pump support materials for all operating PWR plants
and those in the later stages of operation license
review. This reassessment included review of thematerials utilized in the support of 38 potentially affected PWRs. Based on the consultant's evaluation, it
was determined that there are 21 plants whose supports are of questionable toughness and, accordingly,
further detailed plant-specific review is required. Thisdecision concluded the generic study of this subject
under Task A-I2. During the plant-specific reviews
that will follow, either the structural integrity of the
supports must be demonstrated, or measures to assure
their structural integrity will be required.
A report describing the NRC staffs safety evaluation and conclusions and describing its plans for implementation (Le., the more detailed plant-specificreviews referred to above) was issued for comment inNovember 1979. It is entitled, "Potential for Low
Fracture Toughness for Lamellar Tearing in PWR

77

STEAM GENERATORCOMPARTMENTBUMPER BLOCKSTRUCTURALSTEEL COLUMNS

NRC's Generic Safety Issues Task A·12 deals with the ability ofcertain PWR reactor coolant system component supports to with·stand accidents. Shown here are typical reactor coolant pump andsteam generator support structures at the Prairie Island Nuclear
Generating Plant in Minnesota.

Steam Generator and Reactor Coolant Pump Supports," dated September 1979 (NUREG-0577). Review
guidelines and acceptance criteria for use in licensing
reviews of new facilities are also being prepared to incorporate the results of Task A-12 into the staffs
Standard Review Plan. A contractor has been chosen
to undertake the plant-specific reviews of operating
plants during the implementation phase.

The results of an extensive survey by the staffs consultant revealed that, although lamellar tearing is acommon occurrence in structural steel construction,
virtually no inservice failures due to lamellar tearing
are known. Nonetheless, additional NRC-sponsoredresearch is being planned to provide a more definitiveand complete evaluation of the importance of lamellar
tearing to the structural integrity of nuclear power
plant support systems. This research will be a followon effort to Generic Task A-12.

Lamellar tearing is a cracking phenomenon which
occurs beneath welds and is principally found in rolled
steel plate fabrications. The tearing always lies within
the parent plate, often outside the transformed (visible) heat-affected zone (HAZ) and is generally parallelto the weld fusion boundary. Lamellar tearing occurs
at certain critical joints usually within large welded
structures involving a high degree of stiffness andrestraint. Restraint may be defined as a restriction of
the movement of the various joining components thatwould normally occur as a result of expansion and contraction of weld and metal and adj acent regions during welding.

Systems Interactions in Nuclear Power Plants
In November 1974, the Advisory Committee on
Reactor Safeguards requested that the staff give attention to the evaluation of safety systems from a multidiSCiplinary point of view, in order to identify potentially undesirable interactions between plant systems.
The concern arises because the design and analysis ofsystems is frequently assigned to teams with functional
engineering specialties such as civil, electrical,
mechanical, or nuclear. The question is whether the

78

work of these functional specialists is sufficiently integrated in their design and analysis activities toenable them to identify adverse interactions between
and among systems. Such adverse events might occur,for example, because designers did not assure that
redundancy and independence of safety systems were
provided under all conditions of operation required,
which might happen if the functional teams were notadequately coordinated.
The NRC staff believes that its current review procedures and safety criteria provide reasonableassurance that an acceptable level of redundancy andindependence is provided for systems that are required
for safety. Nonetheless, in mid-1977, this task (Task
A-17) was initiated to confirm that present procedures
adequately take into account the potential forundesirable interactions between and among systems.
Because systems interactions are potentially of greatsignificance to plant safety, this issue has been identified as an "Unresolved Safety Issue."
The NRC staffs current review procedures assign
primary responsibility for review of various technicalareas and safety systems to specific organizational
units and assign secondary responsibility to other units
where there is a functional or interdisciplinary relationship. Designers follow somewhat similar procedures and provide for interdisciplinary reviews andanalyses of systems. Task A-17 will provide an independent investigation of safety functions-and

systems required to perform these functions-in order
to assess the degree to which the current review procedures take potential systems interactions into account. This investigation is being conducted by Sandia
Laboratories under contract assistance to the NRC.
The contractor effort, Phase I of the task, began in
May 1978 and is scheduled to be completed in March
1980. The Phase I investigation is structured to identify areas where interactions are possible between andamong systems and have the potential of negating or
seriously degrading the performance of safety functions. The investigation will then identify how NRC
review procedures account for these interactions.
The functional investigation of Phase I is being conducted by the method of "fault tree" analysis. As of theend of the report period, the detailed fault trees werecompleted. Analysis of the fault trees, and the comparison of the results against the NRC review procedures was scheduled to be completed by the contractor by December 31, 1979. A final contractor report
was scheduled to be issued in March 1980. A reportproviding the NRC staffs conclusions based on thecontractor's work was scheduled to be issued in April
1980.
The Three Mile Island Unit 2 accident caused the
NRC staff to consider reorienting the Task A-17 Phase
I effort so as to include improved treatment of suchmatters as operator actions, design errors, andmaintenance procedures. It was decided not to disrupt

For the purpose of its Generic Issues Task A-17, "Systems Interactions," NRC has defined a system as "a set of components or subsystems working as a unit to execute a specific function," Systems
interaction is defined as "a situation where the likelihood of anundesired event is increased due to the relationship between components." The methodology used to address Phase lof this task isreflected in this flow chart.

79

the Phase I effort, which is nearing completion, butrather to consider expanding the Phase II effort to include treatment of TMI-2 related issues. Scoping of thefollow-on work of Phase II is expected to be completed
in April 1980.

Environmental Qualification of
Safety-Related Electrical Equipment
Safety systems are installed at nuclear plants tomitigate the consequences of postulated accidents.
Certain of these postulated accidents could create
severe environmental conditions inside the containment. The most serious of these accidents would be ahigh energy pipe break in the reactor coolant systempiping or in a main steam line. In either case, therelease of hot pressurized water and steam to the con~tainment would create a high temperature environment (250 to 400°F) at high humidity (including
steam) and pressure (as high as 50 psig). For some applications, fission product removal chemicals are added to the containment sprays that are used to reduce
the pressure in the containment. Additionally, someelectrical equipment is predicted to be submerged
following a large pipe break.
In order to assure that electrical equipment in safetysystems will perform its function under accident conditions, the NRC requires that such equipment-principally equipment associated with the reactor trip
system, the emergency core cooling system, and thecontainment isolation and cleanup systems-bequalified to perform in the environment associatedwith the accident. Although such requirements havebeen applied to varying degrees since the early days ofcommercial nuclear power, they have come to be
defined in clearer detail over the years.
The process of clarifying the criteria has given rise tocertain questions regarding the adequacy of qualification tests or analyses. Generic Task A-24, "Environmental Qualification of Safety-Related Electrical
Equipment," was established to address this question
for newer plants.
"IEEE Standard for Qualifying Class IE Equipment
for Nuclear Power Generating Stations," (IEEE std.
323) and its ancillary standards have provided the
focal point for the development of environmental
qualification requirements in recent years. These standards set forth basic requirements for environmental
qualification of electrical equipment and provide
varying degrees of detail for implementation of theserequirements.
The staff requires in part that, for newer plants(specifically those for which a construction permit(CP) safety evaluation report (SER) was issued after
July 1, 1974), the methods and programs developed toqualify safety-related equipment should conform to
the requirements of IEEE 323-1974 and that thisstandard be used as a guide in evaluating thesequalification programs. For plants for which a construction permit SER was issued prior to July 1, 1974,
the staff has required that the qualification programs
be developed in conformance with the guidelines
established in IEEE 323-1971: "IEEE Trial-Use
Standard: General Guide for Qualifying Class IE
Electrical Equipment for Nuclear Power Generating
Stations." This requirement has been applied on a<.:asl'-by-<.:asc basis to oldcr plants that havc been or arc
currently undergoing an operating license OL review.
Several ongoing staff actions related to evaluating theadequacy of environmental qualification of safetyrelated electrical equipment at operating plants are
discussed below under "Other Technical Issues."
Several aspects of equipment qualification are being
pursued at this time by the NRC staff and the nuclear
industry on a generic basis, in order to achieve a more
uniform implementation of requirements established
in IEEE 323-1974. One such activity is a continuing
process of revising and upgrading industry standards
by providing more detailed guidelines for implementing the basic requirements. Another such activity is thedevelopment of NRC staff positions which address
selected areas of the qualification issue. These positions are applicable only to plants that are or will be in
the CP or OL review process and that are required tosatisfy the requirements set forth in either the 1971 or
1974 version of the IEEE-323 standard. This activity
was a part of Generic Task A-24, and was scheduled tobe completed with the publication of interim NRC
positions in NUREG-0588, in December 1979. It is anticipated that a supplemental report will be issued
reflecting any changes to these interim positions whichmight result from investigations of the Three Mile
Island Unit 2 accident, the staffs review of the
responses to Bulletin 79-01 on operating plants, andthe resolution of several issues that are currently being
pursued by the NRC and the nuclear industry such asaging effects, sequential vs. simultaneous testing, etc.
(see Chapter 2).
Further efforts under Generic Task A-24 wereoriginally planned to involve the review of the environmental qualification methods actually used for
qualifying safety-related electrical equipment, pursuant to the requirements of IEEE-Standard
323-1974. The staff had planned to perform thesereviews on a generic basis, rather than on case-by-caselicensing reviews, since this was likely to save time andresources for the NRC and the industry. However, thestaffs initial attempt at performing these genericreviews indicated that it would be unproductive toreview the methods without, at the same time, reviewing the way these methods are used when qualifying
specific pieces of equipment. Therefore, these methods
will be reviewed in conjunction with the review of the80qualification programs submitted as part of theoperating license review of the first plant applications(lead plants) required to comply with IEEE-323-1974.
Since there will be a considerable gap in time between
the issuance of the interim positions and the initiationof the first lead plant review, the scope of Task A-24
has been redefined to eliminate the reviews of specificqualification methods.

Reactor Vessel Pressure Transient
For a number of years, incidents known as "pressuretransients" occurred at various PWR facilities. A
pressure transient occurs when the pressure-temperature limits included in the facility technical specifications for the purpose of protecting the reactor vessel
from brittle fracture have been exceeded. There havebeen over 30 such events. Half of them occurred before
the plant achieved initial criticality (Le., before initiation operation of the reactor); the majority occurred
during startup or shutdown operations. In all of theseincidents, fracture mechanics and fatigue calculations
indicated that the reactor vessels were not damaged
and continued operation of the vessels was acceptable.
Nevertheless, the staff concluded that appropriate
regulatory actions were necessary (1) to reduce the frequency of pressure transient events, and (2) to provide
equipment which would restrict future transients toacceptable pressures. This action was necessary
because reactor vessel safety margins would be reduced over the lifetime of the vessel by neutron irradiation, which reduces material toughness. This matter isdiscussed in more detail above under "Reactor Vessel
Material Toughness."
Upgraded procedural controls were implemented inearly 1977 at operating PWR facilities to reduce thelikelihood of reactor coolant system pressure transients. In addition, system design changes, such asadded pressure relief capability during low
temperature conditions, were also being implemented
in the last two years. No pressure transient events occurred at operating PWRs during the report period.
Task A-26 involved the development of acceptable
criteria for system design changes at operating plants
and for use in the review of construction permit andoperating license applications. All operating PWR
licensees have completed an evaluation of their plant's
reactor coolant system response to potential pressuretransients and, where necessary, have submitted a
description of proposed design changes.
NRC staff review and approval of the proposed
design modifications continues. As of the end of thereport period, 14 facilities have received staff approval. The remaining reviews are expected to be completed by mid-1980.

Residual Heat Removal
Shutdown Requirements
The safe shutdown of a nuclear power plant following an accident not related to a loss-of-coolant accident (LOCA) has been typically interpreted as achieving a "hot-standby" condition (Le., the reactor is shut
down, but system temperature and pressure are still ator near their normal operating values). Considerable
emphasis has been placed on the hot-standby condition of a power plant in the event of an accident or abnormal occurrence. A similar emphasis has been placed on long-term cooling, which is typically achievedby the residual heat removal (RHR) system. The RHR
system can operate only when the reactor coolant
pressure and temperatures have been reduced to
substantially lower than their hot-standby condition
values.
Even though it may generally be considered safe tomaintain a reactor in a hot standby condition for along time, experience shows events sometimes require
eventual cool down and long-term cooling until thereactor coolant system is cold enough to perform inspection and repairs. For this reason, the ability totransfer heat from the reactor to the environment after
a shutdown is an important safety function for both
PWRs and BWRs. It is essential that a power plant beable to go from hot standby to cold shutdown conditions (when this is determined to be the safest course ofaction) under any normal or accident conditions.
This issue was included in the NRC Program for
Resolution of Generic Issues as Generic Task A-31,"RHR Shutdown Requirements." In accordance with
the Task Action Plan for this task, the staffs views on
requirements for residual heat removal systems weretranslated into proposed changes to Standard Review
Plan Section 5.4.7. These proposals were consideredby the NRC's Regulatory Requirements Review Committee (RRRC) during its 71st meeting on January 31,
1978. The RRRC recommended approval of the proposed changes and further recommended that: (1) thechanges be applied on a case-by-case basis to all
operating reactors and all other plants (custom or
standard) for which the issuance of the operating
license is expected before Jannary 1, 1979, and (2) thechanges be backfitted to all plants (custom or standard) for which construction permit or preliminary
design approval applications were docketed before
January 1, 1978, and for which the operating license
issuance is expected after January 1, 1979. These
recommendations were approved by the Director of
NRR and are being implemented. Accordingly,
Generic Task A-31 has been completed.
Standard Review Plan Section 5.4.7 has been
modified to incorporate the approved changes which
are presently being implemented through CP and OL
reviews.

====================================================81
The staff positions on design requirements forresidual heat removal systems were incorporated into
Regulatory Guide 1.139, "Guidance for Residual Heat
Removal," which was issued for public comment in
May 1978. Comments were received during the latterpart of 1978. Work on revision of the guide has been
delayed but it is expected that it will be issued in final
form by late 1980.
In addition, the staff has been reviewing 15
operating nuclear power plants to determine how wellthey meet the recommendations of Regulatory Guide
1.139. Eleven of these plants are included in the
Systematic Evaluation Program (SEP). The remaining
four plants (one representative plant for each of thefour reactor vendors) are of more recent design thanthose included in the SEP. The review for the 11 plants
in the SEP has been completed. Because of possible
design and operational changes based on actions taken
as a result of the Three Mile Island Unit 2 accident, thereview of the other four plants has been postponed.
These reviews are expected to be resumed and completed during calendar year 1980. On the basis of thisreview of representative plants, recommendations for
implementation of Regulatory Guide 1.139 for all
operating plants (except those included in the SEP)
will be presented to NRC management for approval incalendar year 1980. Implementation of SEP findings,
including those related to Regulatory Guide 1.139,
will begin after the end of the Systematic Evaluation
Program, scheduled for May 1982 as of the end of thereport period.

Operating facilities use a variety of design and administrative measures to minimize the potential for
dropping a heavy object over the reactor core or overthe spent fuel pool. These design and administrative
measures have been effective, since no heavy load
handling accidents resulting in damaged fuel rods
have occurred in over 400 reactor-years of U.S.
operating experience. For facilities that have requested increases in spent fuel pool storage capacity,
the NRC has imposed restrictions that prohibit the
movement of any load over the fuel assemblies in the
spent fuel pool that weighs more than the equivalentweight of one fuel assembly. Also, for those plants
where the review of the cask drop or the crane handling system is not complete, movement of shielded casks
over or near spent fuel has been prohibited. It is the
NRC staffs view that continued operation during
review of this generic issue, in compliance with therestrictions cited, presents no undue risk to the healthand safety of the public.
As noted in the 1978 NRC Annual Report, licensees
were requested to examine their current procedures

Control of Heavy Loads Near Spent Fuel
Overhead cranes are used to lift heavy objects,
sometimes in the vicinity of spent fuel, in both PWRs
and BWRs. If a heavy object, such as a spent fuel shipping cask or shielding block, were to fall or tip onto
spent fuel in the storage pool, or in the reactor coreduring refueling, and damage the fuel, there could bea release of radioactivity to the environment. If the
dropped object were large and were assumed to drop
on fuel containing a large amount of fission productswith small decay time, calculated offsite doses could
be high and could exceed the siting guideline values in
10 CFR Part 100.
The NRC staffs review of this safety issue has been
incorporated in the NRC Program for Resolution ofGeneric Issues as Generic Task A-36. The objective ofthe task is to develop standard review criteria which
will reduce the possibility that heavy loads mightcause unacceptable damage to spent fuel in a storage
pool or in the reactor core. The review included a
detailed survey of design provisions and procedures
currently used at operating plants and reevaluation ofcurrent NRC requirements.

This cutaway drawing Ulustrates a typical BWR reactor building
layout, showing the overhead handling system used for movement of
the spent fuel shipping cask and for removal and reinstallation of
the reactor vessel head, moisture separators and steam dryers. Safe
handling of such heavy loads is important to prevent damage toequipment or fuel. Task A-36 will result in upgraded NRC criteria toassure safe handling of heavy loads.

82

for the movement of heavy loads over spent fuel toassure that the potential for a handling accident that
could result in damage of spent fuel is minimized
while the generic evaluation proceeds. In addition, thelicensees were requested to provide information onload handling operations for use in the Task A~36
review. Responses were received from all licensees by
December 1978.
The staff has completed its survey of load handling
operations at operating plants, including design andprocedural measures that prevent or mitigate the consequences of a heavy load handling accident and has
prepared a draft report containing the NRC staffs
resolution of this issue including revised criteria andother recommendations. This report is expected to beissued for public comment in January 1980. The reportwill provide the basis for revisions to the Standard
Review Plan (SRP) and Regulatory Guides, if needed,that can be used in future reviews of new plants andwill provide the basis for implementing additional requirements and procedures,in operating plants.
Although Task A-36 will result in generic criteria,
implementation of these criteria will be dependent onplant design characteristics and the specific proceduresin effect at each particular plant, and will consequently require a plant-by-plant review.

Seismic Design Criteria
NRC regulations require that nuclear power plantstructures, systems and components important to safety be designed to withstand the effects of natural
phenomena such as earthquakes. Detailed requirements and guidance regarding the seismic design of
nuclear plants is provided in the NRC regulations andin Regulatory Guides. However, there are a number ofplants with construction permits and operating
licenses issued before the NRC's current regulations
and regulatory guidance were in place. For thisreason, rereviews of the seismic design of various
plants are being undertaken (principally as part of theCommission's Systematic Evaluation Program) toassure that these plants do not present an undue risk to
the public.
The NRC staff is conducting Generic Task A-40 aspart of the NRC Program for Resolution of GenericIssues. Task A-40 is a compendium of short-term efforts to support the reevaluation of the seismic designof operating reactors, and to support licensing activity
in general. The objective of the task is, in part, to investigate selected areas of the seismic design sequence
to determine the conservatism for all types of sites, toinvestigate alternative approaches to part of the design
sequence, and to estimate quantitatively the overall
conservatism of the design sequence. In this manner
the program will aid the NRC staff in performing its
reviews of the seismic design of operating reactors.

The NRC Office of Nuclear Regulatory Research isalso undertaking a related, but more comprehensive
and long-term program to develop mathematicalmodels to realistically predict the probability ofradiactive releases from seismically induced events innuclear power plants. This Seismic Safety Margin
Research Program will utilize input from Task A-40 in
a number of areas.
Generic Task A~40 is subdivided into two phases.
Phase I includes a number of subtasks related to the
response of structures, systems, and components toearthquakes. These subtasks include studies on: (1)quantifying conservatisms in seismic deSign, (2)electro-plastic seismic analysis methods, (3) sitespecific response spectra, (4) nonlinear structuraldynamic analysis procedures, and (5) soil structure interaction. These studies were performed under NRCsponsored contracts and all were completed by October 1979. Review of the results of these studies isunderway. The results will support the effort on seismic reevaluation of operating plants, particularly inthe area of site-specific definition of seismic input. As
of January 1, 1980, Phase I was scheduled to be completed in February 1980, with the issuance of recommendations for changes in the Standard Review Plan
and Regulatory Guides in those seismic design areas
related to response of structures, systems, and comPOlWllts to seismically induced e\'(~nts.
Phase II of Task A-40 includes several subtasks
related to numerical modeling of earthquake motion
at the source, analysis of near source ground motion,
and attenuation of high-frequency ground motion.
Studies under these subtasks being conducted by NRC
contractors are scheduled for completion by the end of
1980. Review and implementation of the results ofthese studies in terms of recommended revisions to theStandard Review Plan and Regulatory Guides arescheduled for March 1981.

Pipe Cracks at Boiling Water Reactors
Pipe cracking has occurred in the heat affected,
zones of welds in primary system piping in BWRs since
the mid-1960s. These cracks have occurred mainly inType 304 stainless steel that is being used in mostoperating BWRs. The major problem is recognized to
be intergranular stress corrosion cracking (IGSCC) of
austenitic stainless steel components that have been
made susceptible to this failure mode by being "sensitized," either by welding or by post-weld heat treatment. Although the likelihood is extremely low thatIGSCC-induced cracks will propagate far enough tocreate a significant hazard to the public, the occurrence of such cracks is undesirable and measures tominimize IGSCC in BWR piping systems are indicatedto improve overall plant reliability.

====================================================83"Safe ends" (short transition pieces between vessel
nozzles and the piping) that have been highly sensitized by furnace heat treatment while attached to vessels
during fabrication were found to be susceptible to IG~
SCC in the late 1960s. Because they were susceptible
to cracking, the Atomic Energy Commission took theposition in 1969 that furnace-sensitized safe ends inolder plants should be removed or clad with a protec~
tive material, and there are only a few BWRs that still
have furnace-sensitized safe ends in use. Most of these,
moreover, are in small diameter lines and are sub~jected to augmented inservice inspection.
Earlier reported cracks (prior to 1975) occurred
primarily in 4-inch diameter recirculation loop bypass
lines and in 10~inch diameter core spray lines. Morerecently cracks were discovered in recirculation riser
piping (12- to 14-inch) in all foreign plants. All thesecrack locations are part of the reactor primary system.
Cracking is most often detected during inservice inspection using ultrasonic testing techniques. Some pip~
ing cracks have been discovered as a result of small
primary coolant leaks.
In response to these occurrences of BWR primarysystem cracking, a number of remedial actions were
undertaken by the NRC. These actions included:• Issuance of Regulatory Guide 1.44 on "Control of
the Use of Sensitized Stainless Steel."• Issuance of Regulatory Guide 1.45 on "Reactor
Coolant Pressure Boundary Leakage Detection
Systems."• Closely following the incidence of cracking in
BWRs, including foreign experience.
• Encouraging replacement of furnace sensitizedsafe ends.
• Requiring augmented inservice inspection of lines
having less corrosion resistant stainless steel,
especially those that have a high potential forcracking (service sensitive lines).
• Requiring upgrading of leakage detection
systems.
More recently pipe cracking and furnace-sensitizedsafe end cracking have been reported in larger(24-inch diameter) lines in aGE-designed BWR in
Germany with over 10 years of service. Because thesafe ends in that facility had been furnace-sensitized
during fabrication, IGSCC was suspected. As a resultof concerns regarding these furnace-sensitized safe
ends, a safe end was removed and subjected to destructive examination. During laboratory examination of
the re"moved safe end, including a small section of attached pipe, cracks were discovered at various loca~
tions in the safe end and in the weld heat affected zoneof the pipe. The cracks in the pipe weld area were very
shallow with the maximum depth less than 5 mm(about liB-inch) in a wall thickness of about 1.5
inches. Cracking in the furnace-sensitized safe end,
also having a wall thickness of about 1.5 inches, was

REACTORVESSEL

SAFE END AREA WHERE
CRACKING OCCURRED~IENC~~~~.L~I~~~T~I~~R RECIRCULATION OUTLET

H-~..........

I'

RECIRCULATION INLET -;-I-I-~-------i*"(22" DIAMETER)DISCHARGE SHUTOFF VALVE -~~r-+---~SUCTION SHUTOFF VALVEBYPASS VALVE -~..,...-==::r-~

Cracks appeared at one facility in the so~called "safe ends" betw()cn pipes and vessel nozzles. This diagram shows the generallocation of the safe-end area (above), and detail of the cracking
area (below).

10" DIAMETER STAINLESSSTEEL PIPING

THERMAL SLEEVE
AREA WHERE
CRACKING OCCURREDNOZZLE OF REACTORVESSELINNER
VESSELWALL

CARBONSTEEL

84

somewhat deeper. The German experience was thefirst known occurrence of IGSCC in pipes as large as
24 inches in diameter.
In June 1978, a through-wall crack was discoveredin an Inconel recirculation riser safe end (lO-inch
diameter) at the Duane Arnold facility. The crack has
been attributed to IGSCC, although the material inthis instance is different from the Type 304 stainlesssteel that has been historically found to be susceptible
to IGSCC. Prior to safe end removal, ultrasonic examination showed several indications of possible
cracks. Following their removal, cracking was
discovered in all eight safe ends. The cracking appeared to have originated in a tight crevice between
the inside wall of the safe end and the internal thermal
sleeve attachment. Such crevices are know to enhance
IGSCC. Differences in materials, geometry, stresslevels, and crevices appear to make the problem at
Duane Arnold unique to a particular type of recirculation riser safe end (Type I). As a result of this event,ultrasonic examination of the other Type I safe ends in
U.S. BWRs (i.e., at the Brunswick 1 and 2 facility) was
conducted. No significant indications of possible
cracks were found in Unit 2 and one indication was
identified at Unit 1. Although this latter indication
was relatively minor and too small to be reportablepursuant to the NRC Regulations, periodic reevaluation of the Unit was deemed necessary. This ultrasonicindication at Brunswick Unit 1 was remeasured andreevaluated in the presence of NRC ultrasonic testing
consultants at another plant shutdown in January
1979. It was concluded that: (1) there is no apparentchange of this indication between inspections, and (2)although the existence of a very small localized area ofcracking cannot positively be ruled out, the most likely
cause of this indication is irregularities at the weld-tobase metal interface of the first bead weld at the thermal sleeve to safe end weld. This indication will bereexamined during the next refueling outage.
General Electric (the reactor vendor) has been asked
to provide an in-depth report on the significance of recent events, including current inspection, repair, andreplacement programs. They were also asked to address any new safety concerns related to the occurrence of cracking in large main recirculation piping.
Based on information presented by General Electric todate and on extensive staff evaluation, the staff concluded that the recent occurrences do not constitute abasis for immediate concern about plant safety, nor require any new immediate actions by licensees.
Based on the earlier incidents of pipe cracking discussed above, the NRC formed a Pipe Crack Study
Group to: (a) investigate the cause of cracks, (b) makeinterim recommendations for operating plants, and (c)recommend corrective actions to be taken for future

plants. The Study Group published its report(NUREG-75/067) in October 1975, containing recommendations to reduce the incidence of IGSCC in sensitized stainless steel piping. Following staff review of
the Study Group's recommendations, the staff issued
an implementation document (NUREG-0313) which
established staff positions consistent with the recommendations of the Study Group.
As a result of the more recent incidents, the NRC reestablished a second Pipe Crack Study Group on September 14, 1978. The new Study Group specificallyaddressed the following issues:• The significance of the cracks discovered in large
diameter pipes relative to the conclusions andrecommendations set forth in the referenced
report and its implementation document,
NUREG-0313.
• Resolution of concerns raised over the ability ofultrasonic techniques to detect cracks inaustenitic stainless steel.
• The significance of the cracks found in large
diameter sensitized safe ends, and any recommendations regarding the current NRC program for
dealing with this matter.
• The potential for stress corrosion cracking in
PWRs.
• The significance of the safe end cracking at
Duane Arnold relative to similar material anddesign aspects at other facilities.
The new Study Group completed its evaluation in
February 1979 and issued a report, "Investigation and
Evaluation of Stress-Corrosion Cracking in Piping ofLight Water Reactor Plants" (NUREG-0531). The
new Study Group not only reaffirmed the conclusions
and recommendations reached by the previous group
in NUREG-75/067, but also presented some new ideas
to reduce the potential for IGSCC and addressed thesubject of IGSCC in safe ends. On March 13, 1979,
NRC issued a Notice in the Federal Register soliciting
public comments on NUREG-0531. After expiration of
the public comment period and review of the Study
Group's conclusions and recommendations, the staff
initiated Task A-42. The work to be performed under
Task A-42 was defined at that time as the developmentof an update to the implementation document,
NUREG-0313, to incorporate the new Study Group's
conclusions and recommendations and public comments received on NUREG-0531.
Revision 1 to NUREG-0313 was issued in October
1979, and public comments have been solicited on thereport. Revision 1 sets forth the NRC staffs revised
guidelines for reducing the IGSCC susceptibility of
BWR piping. The guidelines describe a number of
preventive and corrective measures acceptable to the
NRC, including guidelines for: (1) corrosion resistantmaterials for installation in BWR piping, (2) methods

85of testing, (3) processing techniques, (4) augmented in·service inspection, and (5) leak detection. The report
also included recommendations for developmental
work to provide future improvements in limiting theextent of IGSCC or detecting it when it occurs.

Containment Emergency Sump Reliability
Following a postulated loss-of-coolant accident,i.e., a break in the reactor coolant system piping, thewater flowing from the break would be collected in
the emergency sump at the low point in the containment. This water would later be recirculated throughthe reactor system by the emergency core coolingpumps to maintain core cooling. This water would
also be circulated through the containment spraysystem to remove heat and fission products from thecontainment. Loss of the ability to draw water fromthe emergency sump could therefore disable theemergency core cooling and containment spraysystems. The consequences of the resulting inability tocool the reactor core or the containment atmosphere
could be melting of the core and/or loss of integrity inthe containment.
One potential way the ability to draw water fromthe emergency sump can be lost is from blockage bydebris. A principal source of such debris could be thethermal insulation normally installed on the reactor
coolant system piping. In the event of a piping break,the subsequent violent release of the high pressure
water in the reactor coolant system could rip off theinsulation in the area of the break. The loose insulation material could then be swept into the sump andblock it.
A Task Action Plan was under development inMarch 1979 when the Three Mile Island Unit 2 accident disrupted work on it. As of January 1, 1980, theTask Action Plan was nearing completion. Nonetheless, several technical studies related to sump reliability which were already underway will either be incorporated into Task A-43 or will provide input into
Task A-43 efforts.
A study program investigating PWR vortex
technology has been completed by the Iowa Instituteof Hydraulic Research and a technical report issued. A
summary report of NRC experience with containment
sump testing is being prepared. This summary will be
issued as a NUREG report in 1980. Based on the Iowa
study program and the review of tests, NRR expects todraft interim positions on sump design guidelines andpreoperational test requirements in early 1980.
Criteria for the evaluation of operating containment
sumps will be formulated at about the same time.
Finall y, a program is being sponsored by the
Department of Energy, in cooperation with NRC, to.
aid in resolving this issue as part of their Light Water
Safety Research Program. This is an experimental pro-

NRC staff mcmbcrs travclcd to thc North Anna Power Station
Unit 1 in Virginia to conduct cvaluatIons of the cmergency recirculation sump as part of its work on Generic Issues Task A-43. The
photo at thc top shows projcct pcrsonnel looking down into thc
rccirculation sump arca. Above, the rcactor containment area wasdeliberately f100dcd to pcrmit ohscrvation of flow pattcrns,
blockage of pipes, types of debris. ctc.

86

gram, begun in July 1979 at Alden ResearchLaboratory, Worcester Polytechnic Institute, to study
the hydraulic aspects of containment sump operation.
The program will continue through February 1981.
It is anticipated that this task can be completed in
1982.

Station Blackout
Electrical power for safety systems at nuclear power
plants is supplied by at least two redundant and in·dependent divisions. The systems used to remove
decay heat to cool the reactor core following a reactor
shutdown are included among the safety systems thatmust meet these electric power supply requirements.
Each electrical division for safety systems includes anoffsite alternating current (a.c.) power connection, astandby emergency a.c. power supply (usually one ormore diesel generators), and direct current (d.c.)sources.
The issue of station blackout involves a study ofwhether or not nuclear power plants should bedesigned to accommodate a complete loss of all a.c.
power, Le., a loss of offsite a.c. sources and all onsite
emergency diesel generator sources. Loss of all a.c. foran extended period of time in pressurized water reactors, accompanied by loss of the auxiliary feedwater
pumps (usually one of two redundant pumps is a steam
turbine driven pump that is not dependent on a.c.
power for actuation or operation), could result in an
inability to cool the reactor core, with potentiallyserious consequences. If the auxiliary feedwater
pumps are dependent on a.c. power to function, thena loss of all a. c. power for an extended period could of
itself result in an inability to cool the reactor core.
Although this is a low probability event sequence, it
could be a significant contributor to risk.
Current NRC safety requirements require as aminimum that diverse power drives be provided forthe redundant auxiliary feedwater pumps. As noted
above, this is normally accomplished by utilizing one
or more a.c. power electric motor driven pumps andone or more redundant steam turbine driven pumps.
One concern is the design adequacy of plants licensed
prior to adoption of the current requirements.
The degree of dependence of decay heat removal
systems on a.c. power supplies and their reliability
with a total loss of a.c. power has recently beenreviewed for a large number of plants. For someplants, modification in design and/or operating procedures were recommended in the short term. Thisevaluation was carried out using simplified analyticaltechniques.
A Task Action Plan for Task A-44 was underdevelopment in March of 1979 when the Three Mile
Island Unit 2 accident disrupted work on this task. As
of January 1, 1980, the Task Action Plan was still

under development. It is anticipated that the task
can be completed in 1982.
Under Task A-44 a more detailed and comprehensive assessment will be performed for both PWRs andBWRs. Preliminary scoping work indicates that thisshould include consideration of: (1) the failure modes
that can result in a station blackout, (2) the probability and frequency of occurence of a blackout including site variability and time dependence, (3) the
potential consequences of a blackout, and (4) potential
preventative and mitigating actions. The results of thiseffort will be used to determine if changes to licensing
criteria are necessary and, if so, to develop criteria foruse in the review of CP and OL applications and forevaluating operating plants.

OTHER TECHNICAL ISSUES
Design Errors in Control Building
In the spring of 1978, Portland General ElectricCompany (PGE), operator of the Trojan Nuclear
Plant, reported design errors in the control building
walls, i.e., conditions at variance with the design
criteria set forth in the Final Safety Analysis Report for
the facility and incorporated into its operating license.
A detailed NRC staff review of PGE's investigation
and analysis of the design revealed the following errors:• The steel reinforcement in the reinforced concretecore of the walls was permitted to be generally
discontinuous and, therefore, the concrete core
could not be relied upon to resist shear (in case ofan earthquake) to the extent assumed in the approved design.
• The shear capacity of the reinforced concrete and. grouted masonry block was not correctly computed.
• The steel reinforcement needed to resist shear
beyond the capacity of the concrete and grouted
masonry block was computed incorrectly,
resulting in a lower level of conservatism than intended.
A detailed reevaluation of the control building in its
existing configuration was performed by PGE to assess
the capability of the structure to withstand the
Operating Basis Earthquake and the Safe Shutdown
Earthquake. The NRC staff determined that there had
been a significant reduction in conservatism anddesign margins, with respect to the control building
seismic capability, below the level intended anddesired for the 33 years remaining in the expected
plant life and that the margins should be appropriately
restored by plant modification. PGE indicated its intent to make such modifications.

87

The NRC staff also determined that, despite thedesign errors, the affected structures remained
qualified to withstand the licensed safe shutdown
earthquake and that there was adequate assurance ofsafety to permit continued operation in the interim
without endangering the health and safety of the
public, provided that no modifications were made thatwould in any way reduce the strength of the existing
shear walls. The NRC staff also concluded that, since
the aBE capability had been reduced, actions thatwould otherwise be required for a O.15g earthquake
would have to be taken in the event of an O.llg earthquake.
On May 26, 1978, the NRC issued an Order dealing
with this matter. The Order called for:• Design modifications to restore the seismic design
margins originally intended to the 'control
building.

• An implementation schedule, to be reviewed andapproved by the NRC.
• Detailed design information for NRC staff review
and approval, together with supporting analyses
and application for license amendments asnecessary to implement these modifications.
• Conditional license waiver of the areas of nonconformance noted above until the control
building has been brought into substantial compliance in these areas. The conditions spelled out
were that no modifications affecting the strengthof the control building shear walls were to bemade without NRC approval and the facility
should be brought to cold shutdown in the eventof a O.llg earthquake at the site and that subsequent restart would require prior NRC approval.
Numerous requests tor a hearing were received, anda hearing was ordered by an Atomic Safety and Licen-

The Trojan Nuclear Plant, situated on the Oregon side of the Columbia River, has been the subject of an NRC staff review concerning the ability of its control building to withstand earthquakes.
Based on an NRC order, Portland General Electric Company(PGE), Trojan owners, submitted plans to modify the building
during 1979, 'and a hearing was set for April 1980. PGE had
notified NRC of the deficiencies in the spring of 1978.

88

sing Board (ASLB). Accordingly, a hearing on interim
operations was held October 23 to November 3, andDecember 11 to 14, 1978. On December 21, the ASLB
issued a Partial Initial Decision that authorized interim operation, with conditions, pending further
hearings on the nature of modifications to the control
building to bring it into substantial compliance withthe requirements of the operating license. The conditions prohibited any modification that would reduce
the strength of existing shear walls; required plant
shutdown in the event an earthquake exceeding 0.08g
should occur at the site; and required modifications tosome pipe supports and restraints prior to resuming
operation in order to ensure qualification of related
piping systems to earthquake levels up to the Safe
Shutdown Earthquake (0.25g).
A conforming amendment was issued on December
22, and plant operation resumed on December 30,
1978.
On January 17,1979, PGE filed a report describing
the proposed modifications to the control building.
These changes consist of the addition of three shear
walls in the existing railroad bay, and the strengthening of the west shear wall by the addition of three-inch
thick steel plate. The design report submitted by thecompany proved to be preliminary in nature and incomplete in some respects. Because of this, extensive
staff questions seeking design details, supporting
analyses and justification for design assumptions were
submitted to the licensee in March, May, July andAugust of 1979. Less than adequate responses to someof these questions caused a delay in issuance of thestaffs Safety Evaluation Report on the modification(scheduled for issuance in September 1979) andpostponement of hearings on the modification which
had been scheduled for October. It also prompted additional staff questions to the licensee in September
and October. The hearing on the acceptability of thesemodifications has been scheduled to commence on
April 1, 1980.
While not directly connected with the original control building design deficiencies described above,
another problem in masonry wall design was identified by PGE in their Licensee Event Report 79-15 ofNovember 4, 1979. Some walls were found to have inadequate structural strength to sustain support reactions from attached piping. This led to concerns overthe load-carrying capability not only of the wall supporting significant piping loads but also of other walls,
and other elements supporting piping and equipment.
A detailed investigation of all masonry walls and otherstructural elements was conducted, which resulted inmodifications to 127 piping supports attached to block
walls and several modifications to other structuralelements. These corrective actions were completed on
December 27, 1979, and NRC review of the matter
was concluded on December 31, 1979.

PWR Feedwater Line Cracks
On May 20, 1979, Indiana and Michigan Power Service informed the NRC of cracking in two feed waterlines at the D. C. Cook Unit 2. Leaking circumferential cracks were identified in the 16-inch lines in theimmediate vicinity of the steam generator nozzles.
Subsequent volumetric examination (by radiography)
revealed crack indications at similar locations in all
feedwater lines of both Units 1 and 2. As a result of aletter sent to all PWR licensees and the issuance of
Bulletin 79-13, inspections have disclosed piping
cracks, crack-like indications or fabrication defects requiring repair in the vicinity of the feedwater nozzles
at 16 out of 25 Westinghouse PWR facilities. To date,eight Combustion Engineering (CE) and two Babcock& Wilcox facilities have been inspected. The two
B&W facilities were found to be free of cracks in the
feedwater piping to steam generator nozzle weld
vicinity. Cracks were found on two of the eight CE
facilities inspected.
The mode of failure at the majority of facilities
where cracking has occurred has been identified as
fatigue assisted by corrosion. The principal cracking atthese facilities has been located in counter-bored areasof the piping at which stress risers, caused by machining to obtain fit-up for welding, are present. Veryshallow cracks have also been identified in the nozzle
at several units.
Repairs at the affected facilities have been made inaccord with improved design and fabrication practices. In addition, the NRC staff has required that
preservice and augmented inspections be performed.
The NRC staff concluded that these measures wereadequate pending the outcome of test programs in progress. The staff has advised licensees that otherremedial measures may be required at a later date, if
results from test programs and/or from further evaluations by the staff show them to be warranted.
Reanalyses of normal piping system stresses andvisual inspections of the feedwater lines have not, todate, uncovered any anomolies that would be expected
to cause cracking. No significant deviations from proper feedwater chemistry control have been discovered.
Thermal stresses, both high- and low-cycle, which
could occur because of the mixing of hot and cold
water in the nozzle region during hot functional
testing, hot standby, start-up and shutdown aresuspected as major factors contributing to the incidence of cracking. Test results from the instrumentation placed on the affected plants have shown that
thermal stratification does exist at start-up, hotstandby and shutdown conditions. A Utilities Owner'sGroup has been formed to conduct test programs toidentify the cause or causes of cracking and to find along-term solution to the problem.

=============================================89
The NRC has instituted an Action Plan regarding
the PWR Feedwater Cracking problem. The following
items are included in the Action Plan:(1) Consequence of Cracks-This item included
identification of possible challenges to pipe
integrity, and consideration of system effects interms of the health and safety of the public.
(2) Failure Investigation: Cause of Cracking-This
item included (a) independent metallurgicalanalyses to identify the mode of failure; and (b)review and independent stress analyses to verify
the licensees' concl usions.
(3) Review PWR Designs and Operations-This
item includes review of the piping layout,
fabrication and inspection history, preservice
and inservice performed, and evaluation of theoperating histories.
(4) Evaluation of the Remedial and Corrective
Actions- This item includes review of therepairs effected; the evaluation of the efficacy of
the repairs in light of the results of the test
results from the Westinghouse PWR facilities;
and evaluation of design changes and operatingprocedures as a long term solution to the problem.
The NRC staff has considered the safety significanceof the feedwater piping cracks and has concluded,
based on the available information, that the most
severely degraded piping found to date (D. C. Cook
Unit 2) would be unlikely to rupture in the event of anearthquake, though it might leak. The staff considers
it conceivable that a severely degraded feedwater line
may rupture from a severe water hammer event.
However, it is considered unlikely that a severe water
hammer event would occur in two or more feedwaterlines simultaneously. Thus, the worst consequence tobe reasonably expected in a facility with degradedfeedwater piping which experienced a severe water
hammer would be the rupture of a single feedwater
line. Because this event has been considered as a design.
basis accident, the facilities are designed to protect
against the occurrence and to contain and control its
consequences.

Mechanical Operability ofContainment Purge Valves
In November 1978, the NRC requested that all
licensees of operating reactors respond to generic concern about containment purging or venting during
normal plant operation. The generic concerns
involved both electrical and mechanical aspects ofcontainment purge valve operability. First, events had
occurred in which licensees overrode or bypassed thesa,ftey actuation isolation signals to the containment

isolation va1ves. These events were determined to beabnormal occurrences and reported to Congress inJanuary 1979. Second, recent licensing reviews haverequired a demonstration, by test or analyses, that
containment purge or vent valves would shut without
degrading containment integrity during the dynamic
loading imposed by a postulated design accident, a lossof coolant accident (LOCA).
The November 1978 request emphasized the importance of the mechanical qualification of purge andvent valves. For quick closing capability most facilities
use butterfly valves for containment isolation.
The concern centered around the capability of thecontainment isolation valves in the purge and ventsystems-after being opened during hot standby, hot
shutdown, startup or power operation modes-to havethe capability to close against the fluid dynamic conditions of a LOCA upon receipt of an isolation signal.
These fluid or aerodynamic forces originate from thepressure drop imposed across the closing valves by theascending pressure in containment following a LOCA.
Typically these valves will receive an isolation signal
to close either from high radiation monitors, highpressure monitors, or low reactor water level for
BWRs.
Potential failures affecting the purge and ventvalves could lead to degradation in the cO'ntainment
integrity and, for PWKs, a degradation in ECCS performance.
From staff studies and discussions with manufacturers involved in supplying these valves, the following
conclusions can be made:(1) Most valves of this type will tend to close under
the dynamic forces of a LOCA.
(2) Partial opening of the valves between 30 0 and
50 0 of full open will in most cases significantly
reduce dynamic loads put on valve components.
(3) Demonstration of operability for most valves ofthis type can be obtained through analysis andprevious testing data.
Guidelines for demonstration of purge and ventvalve operability have been developed and issued to all
licensees and will be used to assess the valves installedin operating plants. This effort is scheduled to be completed !'y mid-fiscal year 1981. In the interim,
licensees have been asked to limit the opening of theirvalves to between 30° and 50° of full-open and to limit
the use of these valves until such time as long term
operability can be demonstrated. It is anticipated thatsome systems modification may be reqUired in attaining these goals.

Environmental Qualification of
Safety-Related Electrical Equipment
At Operating Plants
In the fourth quarter of calendar year 1978, the staff
conducted inspections of the activities of licensees of

90

all operating reactors, in response to a staff issued circular, dated May 31, 1978, concerning the qualification of safety-related electrical equipment. The
inspections disclosed that the review and resolution ofproblem areas encountered by the licensees in thiseffort were not receiving sufficient attention. Also
identified were certain components for whichdocumentation was lacking as to which were foundqualified and which were found unqualified for theirintended service. In order that this subject be given theproper emphasis, the staff issued Bulletin 79-01, dated
February 8, 1979, to licensees of all operating reactors
facilities. The bulletin required written responses fromthe licensees, within 120 days, regarding the results of
their reviews of environmental qualification of all
safety-related electrical equipment. Further, it
required that the licensees report to the staff, within
24 hours of discovery, any equipment determined to
be unqualified for its service conditions. A supplement
to this bulletin, dated June 6, 1979, was also issued
providing feedback to the licensees regarding
deficiencies in the environmental qualification of certain pilot solenoid valves.
In response to the bulletin and its supplement thestaff has received several 4424-hour reports" (23separate reports from 29 different nuclear plants)
involving seven different types of equipment- (1)limit switches mounted on safety-related valve stems
to indicate stem position; (2) containment isolation
valve motor operators; (3) instrument and control
cable insulated terminal lugs; (4) aluminum limit
switch housings on containment isolation valves; (5)ASCO pilot solenoid valves for miscellaneous valve air
operators; (6) terminal blocks enclosed in boxes; and(7) interconnecting wires. In each instance where an
item of equipment was determined to be unqualified,
the staff immediately evaluated the impact on thehealth and safety of the public and the adequacy of theremedial steps proposed by the licensees. In somecases, the licensees elected to replace the unqualified
equipment immediately; in others, a basis for continued operation pending corrective action at a futuredate was provided. In those cases where the licensees
proposed to continue to operate the plant for a periodof time before shutting down and replacing theaffected equipment, the following factors were considered in the staff evaluations of whether the plants
could contiaue to be operated safely: (1) redundant/diverse components available to perform therequired safety functions; (2) locking of the affectedcomponent in its safety position; (3) administrative
actions and revised operating procedures; (4) additional operability tests and inspections; (5) postaccident mitigating actions available; and (6) fail safe
design features. In all cases where continued operation
was requested by the licensees, based on a plantspecific safety evaluation, the staff has concluded(contingent, in some cases, upon additional staff requirements being satisfied) that the plants could continue to be operated safely.
An NRC task group has been formed to review indetail the licensees' responses to Bulletin 79-01. The
reviews will be conducted in accordance withguidelines being prepared specifically for evaluating
the qualification of Class IE equipment in operating
reactors. The guidelines will address all of the significant aspects of the most current industry standard forClass IE electrical equipment qualification, lEE Std.
323-1974.
The Office of Nuclear Reactor Regulation has
prepared, guidelines for the Office of Inspection and
Enforcement to use in identifying safety-related electrical equipment installed in operating reactors whose
documentation does not provide reasonable assurance
of environmental qualification.

Pipe Support Base Plate Problem
During the review of pipe support designs at theNorth Anna Units 1 and 2 and Shoreham Unit 1, a
potential design deficiency was discovered in themodeling assumptions used to calculate pipe support
anchor bolt loads. Inspections at Shoreham also
revealed deficient installations of a large number of
the anchor bolts. During June 1978, the NRC initiatedaudits of engineering firms involved in the design ofpipe supports to determine the potential generic extent
of the problem in operating plants. As a result of theseaudits and the review of operating experience related
to pipe support failures, the NRC issued Bulletin 79-02
on March 8, 1979, to all operating plants and to allplants with construction permits. The bulletin
requires all licensees to re-calculate anchor bolt loads
using appropriate modeling assumptions and to
inspect the anchor bolt installations to insure thatanchorage systems have an adequate margin of safety
for the imposed loads.
Preliminary review of licensee responses has been
completed. Modifications to most plants should becompleted by March 1980. Several operating plants
are making substantial modifications to their existinganchor bolts to meet the requirements of the bulletin.
The staff has reviewed the existing design margins andhas determined that sufficient margin exists withrespect to the functional capability of the affectedsafety systems during a seismic event to allow operation while these modifications are being made.

Nonconformance of Actual Installation of
Safety-Related Piping Systems
To Design Documents
During seismic reanalyses of safety-related pipingwith acceptable computer codes (see "Shutdown and91

Seismic Reanalysis of Five Operating Reactors"
below), inspection of several piping systems by NRC
and its licensees identified substantial differences between design documents and the as-built condition oftht' piping and supports. Some examples are:
• Surry I-Mislocated supports, wrong support
type, and different pipe geometry.
• Fitzpatrick-Supports installed which were notaccounted for in the analyses, cracked welds on
support components, and bent rod hangers.
• Brunswick 1 and 2- Undersized pipe supports.
• St. Lucie l-Missing seismic supports andmislocated supports.
In order to obtain accurate and valid results, thedata used as input to the required pipe stress analyses
and associated support evaluations, e.g., pipe size,
geometry, support location and type, must reflect,
within construction tolerances, the "as-built" condition of the plant. Specific seismic reanalysis results for
Surry 1, Fitzpatrick, and Beaver Valley 1
demonstrated that when the analyses reflected the asbuilt condition of the facilities, piping stresses, nozzle,
penetration, and support loads exceeded their
allowable values. Several support modifications, additions, and deletions were necessary to the facilities'
piping and supports in order to show compliance withthe originally intended design criteria.
The magnitude and type of discrepancies discussed
above were discovered' at 11 power reactors, indicating that the problem may be generic and could
substantially lessen the ability of a large number offacilities to adequately withstand the effects of aseismic event. Therefore, on July 2, 1979, the NRC
issued Bulletin 79-14, addressing seismic analyses for
as-built safety related piping systems. Revision 1,
which l'xempts from the requirements of this bulletin
all two-inch and smaller piping not computer analyzed, was issued July 18, 1979. Additional guidance wasprovided by the issuance of supplements to the bulletin
on August 15 and September 7, 1979.
Bulletin 79-14 requests that all licensees verify thatthe seismic analysis input information accuratelyreflects the facility as actually constructed. Licensees
were requested to inspect piping geometry, piping
support and restraint designs, pipe attachments, andvalve and valve-operator locations and weights. The
bulletin requests that licensees establish an ad hoc
inspection program to be completed within 120 days.
Further, the bulletin requires that licensees resolve
specific nonconformances by either making changes tothe system so that it conforms to design or by correcting the erroneous seismic analyses to demonstrate
conformance of the as-built system to design criteria.
It also requires that licensees take action to correct
administrative problems which could allow this problem to recur.

Because of the extensive effort' required-involving
highly qualified stress analysts and related engineering
disciplines-the time for completion of the requirements of the bulletin has been extended beyond
120 days. Most facilities are scheduled to complete therequirements by April 1980.
The NRC is assessing the results of the continuing inspections by licensees. Although not yet complete, theinspections conducted to date indicate that mostfacilities will be required to make some equipmentchanges. Most of the operating plants have alreadymodified or added piping supports because of deviations found between the existing, as-built, equipment
and design documents. Several plants, including Fort
St. Vrain, Millstone Unit 2, D.C. Cook Unit 1 andRancho Seco, have shut down for various lengths oftime, in compliance with technical specification requirements, as a result of discrepancies discovered
during the inspections.
Several architect-engineer firms and licensees havebeen audited to discuss and examine their proceduresin the implementation of bulletin requirements, andmore audits are planned for the future.

Shutdown and Seismic Reanalysis ofFive Operating Reactors
The Nuclear Regulatory Commission staff ordered
five plants to shut down on March 13, 1979, until
reanalysis and necessary modifications were made tosafety-related piping systems, in order to bring theminto conformance with requirements for withstanding
earthquakes. The plants ordered shutdown were;
Beaver Valley Unit 1, James A. FitzPatrick, Maine
Yankee and Surry Units 1 and 2.
Stone and Webster Engineering, the architectengineer for these five plants, and Duquesne Light
Company, the licensee for the Beaver Valley facility,
had earlier reported to the NRC, during a meeting on
March 8, 1979, that they had discovered that analgebraic summation method has been used to combine seismic forces in the computer~ode called
SHOCK II. The algebraic summation method canresult in cancellation of seismic forces and thus may
result in the prediction of stresses significantly lower
than would be predicted by NRC-approved techniques. Following the meeting on March 8, membersof the NRC staff met for three days with Stone &Webster Engineering officials in Boston. Additional
analyses of piping systems for the Beaver Valley facility were performed. These analyses indicated signifi.
cant overstress in the piping systems under postulated
earthquake conditions when computer codes wereused which did not combine seismic loads algebraically. Piping systems involving the integrity of the reactor
coolant pressure boundary, emergency core coolingsystems and safe shut down systems were involved. It

92

The Virginia Electric Power Co. began repairs on the steam
generator of Surry Power Station, Unit 2, including the repla<.'Cment of the lower sections of all three steam generators with new,improved sections. The effort will take about one year to complete. Shown in this photograph is onc of thc sections being removed through the equipment hatch of the reactor containment. NRC
is arranging to obtain one of the steam generator sections for usc in
its research and development programs.

was also determined that the same computer code
(SHOCK II) was used in the design of four otherfacilities. The NRC staff ordered all five plants shutdown because there was no assurance that a severe
earthquake at any of these facilities would not cause
an aCCident, damage emergency core cooling systems,
and prevent safe shutdown of the plant.
The required reanalysis and necessary modifications
were completed for Maine Yankee and Beaver Valley
and orders were issued permitting resumption ofoperation, on May 24, 1979 and August 8, 1979,
respectively. Sufficient reanalysis and modifications
were completed for FitzPatrick and Surry Unit 1 topermit orders, issued on August 14, 1979, and August24, 1979, respectively, allowing resumption of operation for 60 days while some remaining pipe support
analyses were completed.
Surry Unit 2 was shut down for steam generator
repair and replacement prior to the March 13, 1979,
shutdown order. Because of the long shutdown forsteam generator work, the seismic reanalysis requiredby the order was delayed by the licensee. It was not

anticipated that the required seismic reanalysis would
lengthen the plant shutdown.
Several actions have been taken by the NRC staff
related to review, evaluation and approval of computer codes used for seismic analysis of saftey-related
piping. Following issuance of the show cause orders, acomputer code verification program was initiated by
the staff, with three principal parts: (1) review ofactual computer code listings, (2) solution of NRC
benchmark problems to compare results to known
values, and (3) independent check analyses of piping
problems using NRC's own computer code. In addition, the NRC staff reviewed the development of themathematical model which represents the piping
system.
On April 13, 1979, the Florida Power and Light
Company, licensee for Turkey Point Units 3 and 4,
reported that algebraic summation techniques had
been used by Westinghouse in the design of the main
reactor coolant system piping. The NRC reviewed theresults of Westinghouse's reanalysis, determined thatpiping design was acceptable, and permitted resumption of operation of both Turkey Point units.
However, as a result of this information, an IE
Bulletin was issued on April 14, 1979, requiring all
licensees to review the computer codes used in thedesign of safety-related systems to determine if
algebraic summation had been used. A total of 24 additional plants had used an algebraic summation technique. Four of these plants were still under construction and had not yet been issued operating licenses.
The computer codes identified were:
SHOCK II-Stone & Webster Engineering
WESTDYN- Westinghouse
DAPS-General Electric
PIPDYN II-Franklin Institute
ADLPIPE-Arthur D. Little Company
The NRC staff has required reanalysis of all affectedpiping, with modification as necessary, and computercode verification for those codes used for reanalysis.
The majority of the 25 operating reactors usedalgebraic summation methods on very few piping
systems and had reanalyzed these systems prior to
responding to the bulletin. In a few cases (Pilgrim 1,
Brunswick 1 and 2, Indian Point 3 and Salem 1) theuse of algebraic summation was more extensive. One
unit, Salem 1, shut down since March 31, 1979, forrefueling and other modifications, did not resume
operation until November 13, following resolution of
the algebraic summation issue. All other units havebeen resolved completely or, based upon NRC staff
evaluation, have been permitted to continue operation
during reanalysis. In each case where continued
operation was permitted (Brunswick 1 and 2 andIndian Point 3), the analysis methods used and themargin allowed in the piping design were such thatmodification to piping systems was unlikely. The staff.

===================================================93
however, required detailed reanalyses to confirm thatthe designs were acceptable.

Petition for the Seismic Reanalysis
Of Operating Reactors
On March 28, 1979, the Union of Concerned Scientists (UCS) filed a petition which proposed that the
NRC require all plants with an operating license toperform a seismic reanalysis within a 120-day time
period. The proposed reanalysis consisted of reevaluation of: (1) the magnitude of the safe shutdown earthquakes; (2) the free-field ground motion at the site; (3)the motions of the structures during a seismic event;(4) the plant equipment motions during a seismic
event; (5) the seismic loads on structures, systems andcomponents in appropriate combinations with otherloads, and the corresponding allowable loadings; and(6) the conformance of the "as-built" plant to thedesign specifications. This petition was issued based
upon concerns arising from issues surrounding thefive-plant shutdown, including the lack of rigorous
computer code verifications, and concerns regarding
the evolution of seismic design criteria through theyears.
Currently acceptable seismic design requirements
for nuclear plants are delineated in Title 10 CFR, the
NRC Standard Review Plan, and associatedRegulatory Guides and engineering codes and standards.
There are many variations in the parameters used todefine the ground motion imparted by an earthquake.
In evaluating the seismic hazard to a plant for a given
definition of the ground motion, a detailed engineering evaluation is conducted considering groundmotion, foundation/structure interaction, piping
system equipment, and component response. The
uncertainties in the various steps of the overall analysis
and design lead to conservative assumptions being
made in each step.
The NRC has four programs underway to assess theseismic design adequacy of operating nuclear plants:• Bulletins have been issued to each licensee requiring evaluations and, if needed, hardware
modifications to affected plant systems. These
items to be evaluated include: (1) verification of
the desired safety margins in piping supports anchored by expansion bolts to concrete structures;(2) verification that the seismic analysis of safetyrelated piping systems was based on acceptablesummation methods of seismic loading components; (3) verification that as-installed piping
systems and supports are essentially the same asused in seismic analysis and design.
As the reviews proceed, the NRC will take anynecessary actions deemed appropriate. The

review to date indicates that some installationand design deficiencies exist; these are being
resolved in a timely manner.
• The Systematic Evaluation Program (see discussion later in this chapter) includes review of theseismic design of 11 older operating plants. No
major deficiencies in the seismic design which
would affect public safety have been identified.
Several issues have been identified which will require more detailed studies, and possibly retrofitting, to verify the adequacy of the seismic design
to meet the intent of current design criteria.
• The Seismic Design Criteria study (Task Action
Plan A-40) is a short-term program to provide
generic, quantitative estimates of the conservatisms in selected individual parts of and theoverall seismic design when following current
criteria.
• The Seismic Safety Margins Research Program,
expected to continue for about six years, isoriented toward improvements in seismic design
methodologies. These studies are being carriedout concurrently with and will extend beyond the
Seismic Design Criteria short-term effort.
The NRC response addressing in detail the concerns
voiced in the UCS petition was in preparation at theclose of the report period. (The UCS petition was
denied by the Director of Nuclear Reactor Regulationin January 1980.)

Humboldt Bay(The background to licensing problems associatedwith this facility can be found in the 1977 NRC Annual Report, pp. 26-27, and the 1978 NRC Annual
Report, pp. 49-50.)
The Humboldt Bay Power Plant (Cal.) has been
shut down since July 2, 1976, because of unresolved
geologic/ seismic concerns regarding local geologic
fault definition and the potential for surface faulting
at the site. Since 1976, the Pacific Gas and ElectricCompany (PG&E) licensee for the facility, has conducted extensive geologic investigations and plantseismic modifications.
A request for hearing with respect to resumption ofoperation of the. facility was submitted by representatives of individuals from the Humboldt Bay area.
The NRC Atomic Safety and Licensing Board (ASLB)
has directed that such a hearing be held. On May 7
and June 19, 1979, the ASLB issued separate but
related orders to the licensee requiring that it provide
the board and interested parties progress reports on
the status of ongoing geologic investigations. The
licensee is providing such information bi-monthly andestimates that the earliest practicable date for proceeding with a hearing is October 1, 1980. At the closeof the report period, the ASLB was considering thelicensee's motion to hold the proceeding in abeyance
until October 1980.

94

Possible Faulting Near Reactor
On October 24, 1977, the NRC staff ordered that
the General Electric Test Reactor (GETR) at theVallecitos Nuclear Center in California be shut down.
This action was based on evidence, revealed by field
investigations, of possible faulting (Verona Fault) at
the GETR site. Since surface faulting was not a designbasis for the GETR facility and the appropriate
seismic design ground acceleration was in question,
the licensee was directed to show cause why thesuspension of operation should not be continued.
Following attempts to argue that the Verona Fault
did not exist, the licensee conducted an extensive fieldtrenching investigation. This investigation revealed
additional evidence of faulting at the site. In its reporton these investigations, submitted February 28, 1979,
the licensee concluded that the origin of observed offsets was pr-obably large-scale landsliding but could
also be the result of earthquake faulting. Furthermore,
the licensee proposed a zero surface offset design basis
for the reactor building and no more than three feetdisplacement on observed offsets.
By letter dated September 27, 1979, the NRC staff
issued its evaluation of the seismology and geology of
the GETR site. The NRC staff concluded that a surface offset of two and a half meters could occur
beneath the GETR. This is in excess of the one metersurface offset to which the GETR facility has been
analyzed by the licensee.
This issue will be subject to further reviews by the
NRC Advisory Committee on Reactor Safeguards andthe Atomic Safety and Licensing Board.

Fire Protection
Following the fire at the Brown's Ferry Plant inMarch 1975, the NRC initiated a review of the fire
protection programs for all operating plants and forplants not yet operational. Improved guidelines havebeen developed and are being implemented.
Minimum requirements for specific aspects of fire protection for operating plants are being added to 10 CFR
50. The fire protection program reviews have been
completed for the 70 licensed power plants andmodifications to improve plant capabilities are being
implemented. The modifications to most plants will bemade by late calendar year 1980.,
On November 4, 1977, the Union of Concerned
Scientists (UCS) filed a Petition for Emergency andRemedial Action. Part of ,this petition dealt with fire
protection concerns at plants under construction andat operating plants.
The Commission issued an order on April 13, 1978,
denying the UCS petition on the basis that plants
under construction or in operation are in compliancewith General Design Criterion 3-Fire Protection.

On May 2, 1978, the UCS submitted a petition
requesting that the Commission reconsider its April
13" 1978, decision on the earlier petition filed on
November 4, 1977. The Commission took this petition
under consideration, and was reviewing public andstaff comments, developed as a result of the reconsideration, at the close of the report period.

Control Rod Guide Tube Integrity
In December 1977, extensive wear and some holes
were observed in the upper section of numerous control rod guide tubes at Northeast Nuclear Energy
Company's Millstone Unit 2 facility (Connecticut).
Subsequent inspections at other facilities with reactors
designed by Combustion Engineering (CE) disclosed
similar indications of guide-tube wear. The cause wasfound to be flow-induced vibration of fully withdrawn
control rods. The rod tips, vibrating against the guidetubes, induced degrading wear, probably aided bycorrosion.
The safety significance of the incidents is related to
the functions of the guide tubes. They serve both asfuel assembly structural members and as channels forcontrol rod movement. Thus, a guide-tube failure
could adversely affect either the preservation of acoolable core geometry or the scram (shutdown)
capability of the control rods, or both.
The observed wear of the guide tubes thus far has
been confined to facilities designed by Combustion
Engineering (CE). There are basic differences
between the CE design of the control rod systems,
which insert into the guide tubes of the fuelassemblies, and the other designs (Westinghouse andBabcock & Wilcox). These design differences appear
to have reduced the severity of wear on the guide tubes
in the latter vendors' facilities. However, such wear in
Westinghouse and Babcock & Wilcox plants and in
Exxon Nuclear fuel assemblies is under investigationby the NRC staff.
To overcome the susceptibility to wear of the guidetube material (Zircaloy-4) and to recover the design
margin lost by wear, CE designed stainless steel sleeves
for use in the guide tubes. Prior to installation ofstainless steel sleeves during a refueling outage,
operators of CE reactors instituted the practice ofinserting the control rods three inches further into thecore than the normal fully withdrawn position. That
action both reduced the local wear intensity and provided added assurance of scram capability. NRC
approval was granted for this short term administrative procedure allowing continued operation.
The use of sleeved guide tubes was approved by the
NRC as an interim repair to mitigate the gUide-tube
wear on a cycle-specific basis. In conjunction with theuse of the stainless steel sleeves, the staff required that
inspection programs be submitted for review andapproval well in advance of refueling shutdowns.

=====================================================95Additional out~of~reactor hot loop testing by CE
showed the important role of flow-induced vibrationof the control rods in the guide-tube wear problem.
The vibration and, hence, the wear, was reduced bydecreasing some of the guide-tube coolant (water)
flow. Two fuel assembly modifications were designed
to reduce the coolant flow. One involved inserting a
splined cylinder in the top of the guide tube. The second involved reducing the size and number of flowholes in the bottom of the guide tube. Both modifications, in limited number, are being tested in currently
operating cores to confirm the loop test results.
The NRC has closely followed the analyses andexperiments performed by CE and is in substantial
agreement with the vendor that the results point tocontrol rod flow-induced vibration as the principal
factor in guide-tube wear. Therefore, design modifications intended to alter flow in the guide-tubes were
judged appropriate. The NRC has approved themodified designs for limited operation on the basisthat they will mitigate the wear problem. Approval ofeither design modification as a final solution to the
problem will be contingent upon the results of further
out-of-reactor experiments and examination of the,modified assemblies which are currently subject to inreactor operations.
The first opportunity to evaluate the performance ofthe sleeved guide tubes after reactor operations occurred during the Millstone Unit 2 refueling outage in thespring of 1979. Subsequent to the Millstone 2 refueling, the St. Lucie Unit No.1 (Florida) and the Calvert
Cliffs Unit No.1 (Maryland) also provided evidence
on the performance of the sleeved guide tubes. These
inspections indicate that the sleeving modification has
performed well as an interim solution to mitigate theguide-tube wear but that it does not eliminate thecause of the wear. (During the October-November
1979 refueling outage Calvert Cliffs Unit 2 wasscheduled to undergo inspections of modifications
made as interim solutions to guide tube wear.)
The NRC staff will continue to maintain close
liaison with representatives of the licensees and vendors on this issue and any related problems. Approvals
have been granted to allow operaton of the CE plants
on a cycle-specific basis with the stainless steel inserts.
All proposed programs have been reviewed prior to
taking action at any facility, and the staff has requiredthat all inspection programs continue to be submitted
for review well in advance or refueling shutdowns.

Four procedural options are available (see 1976
NRC Annual Report, p. 36, for details) to applicants
for standardization of nuclear power plants:"Reference Systems" (approved design used repeatedly
by reference), "Duplicate Plants" (approved design forseveral identical plants), "License to Manufacture" (approved design for manufacture of identicalunits at the central location), and "Replicate Plants"(reuse of recently approved custom design).
Since June 1973, when applications were first accepted which included a standardization option, thestandardization program has realized substantial progress. Overall, approximately two-thirds of the applications received in the 1974-1978 time frame have
employed one or more options of the standardization
program. See Table 3 for a listing of the status of applications.
In August 1978, the Commission approved anumber of changes to the program to encourage its expanded use, as well as to incorporate both industry
and regulatory changes introduced since the program
was first announced in 1972. The revised program
adopted a good many such changes, some of which are
as follows:

PROGRESS IN STANDARDIZATION(4) A qualification review was defined for replicate
plants and the period for replication was established as 3 years after publication of the base plant
Safety Evaluation Heport.

The NRC believes that standardization of the designof nuclear power plants is in the interest of publichealth and safety, and of effective and efficient regulation. Thus, the NRC is committed to the support andexpanded use of standardization within the Commission's regulatory activities.

(1) The term of holders of all new preliminary design
approvals (PDAs) for reference system designs was
extended from three to five years. Holders of allissued PDAs were given the opportunity to extend
them to a full 5-year term.
(2) Final design approvals (FDAs) for reference
system designs were eligible for reference in applications for construction permits. Two types of
FDAs were established. The first, denoted FDA-I,
can be referenced from the time it is docketed to 3
years after expiration of the PDA on which it isbased. The second, denoted FDA-2, can bereferenced from the time it is docketed to 5 years
after it is approved.
(3) A qualification review was devised to permit the
duplicated plant concept to be used in a manner
similar to the reference system concept. In thisregard, five-year preliminary duplicate design
approvals (PDDAs) and final duplicate design
approvals (FDDAs) were established which can beused in new applications for construction permits
in a manner similar to the use of PDAs and FDAs
under the reference system concept.

(5) A 5-year period of design approval was established
for manufacturing licenses and an uppper limit of
10 units was established.

96

Table 3. Standardization Applications(as of August 31, 1979)
APPLICANTPROJECT

DOCKET DATECOMMENTS

Reference Systems

Nuclear Island
GESAR-238(NI)General Electric

7/30/73

Nuclear Island, PDA-l(Preliminary Design Approval) issued
12/22175

Turbine Island
C F BRAUN SSAR

C.F. Braun

12/21174

Turbine Island Matched
TO GESSAR-238(NI). PDA-5 Issued
5/07176

Nuclear Steam Supply System (NSSS)
BSAR-205

Babcock & Wilcox

3/01176

PDA-12 issued 5/31/78

BSAR-241

Babcock & Wilcox

5/14/74(withdrawn)

CESSAR

Combustion Engineering

12/19/73

PDA-2 issued 12/31/75

GASSAR

General Atomic

2/05/75

Review suspended at request of applicant.

GESSAR-238

General Electric

10/16/75

PDA·lO issued 3/10/77

GESSAR-251

General Electric

2/14175

PDA-9 issued 3/31177

RESAR-3S

Westinghouse

7/31175

PDA-7 issued 12/30176

RESAR-41

Wesinghouse

3/11174

PDA-3 issued 12/31/75

RESAR-414

Westinghouse

12/30/76

PDA-13 issued 11114178

BOPSSAR/BSAR-205

Fluor Power

10/31177

BOP matched to BSAR-205

BOPSSAR/RESAR-41

Fluor Power

1127176

PDA-11 issued 8/17/77
BOP matched to RESAR-41

ESSAR/BSAR-205

Ebasco5/19178

BOP matched to BSAR-205

ESSAR/CESSAR

Ebasco2/02178

BOP matched to CESSAR

ESSAR/RESAR-414

Ebasco11/23177

BOP matched to RESAR-414

GAISSAR/BSAR-205

Gilbert Commonwealth8/21178

BOP matched to BSAR-205

GAISSAR/CESSAR

Gilbert Commonwealth8/21178

BOP matched to CESSARGAISSAR RESAR-414

Gilbert Commonwealth8/21178

BOP matched to RESAR·414

GIBBSSAR

Gibbs & Hill

5/10177

BOP matched to RESAR-414

SWESSAR/BSAR-205

Stone & Webster

12/22175

BOP matched to BSAR-205

SWESSAR/CESSAR

Stone & Webster

10/21174

BOP matched to CESSAR
PDA-6 issued 8/16176

SWESSAR/RESAR-3S

Stone & Webster

10/02/75

BOP matched to RESAR-3S
BPDA-8 issued 3/31177

SWESSAR/RESAR-41

Stone & Webster

6/28174

BOP matched to RESAR-41
PDA-4 issued 5/05/76

Balance of Plant (BOP)====================================================97
PROJECTAPPLICANTDOCKET DATECOMMENTS

Utility Applications Using Reference Systems
Cherokee 1,2&3

Duke Power5/24174

References CESSAR. CP issued 12/30177

Perkins 1,2&3

Duke Power5/24174

References CESSARSouth Texas 1&2

Houston Light and Power Co.

7/05174

References RESAR-41 CPs issued 12/22175

WPPSS 3&5

Washington Public Power
Supply System

8/02174

References CESSAR
CPs issued 4/11178

Palo Verde 1,2&3

Arizona Public Service

10/07/74

References CESSAR. CPs issued OS/25176

Hartsville 1,2,3&4

Tennessee Valley Authority

11/22174

References GESSAR-238(NI)CPs issued 05/09177

Palo Verde 4&5

Arizona Public Service

03/31178

References CESSARBlack Fox 1&2

Public Service of Oklahoma

12/23175

References GESSAR-238 (NSSS)

Phipps Bend 1&2

Tennessee Valley Authority

11/07175

References GESSAR-38
CPs issued 11 16178 (NI)

Erie 1&2

Ohio Edison Co.

3/01177

References BSAR-205

Yellow Creek 1&2

Tennessee Valley Authority

3/16176

References CESSAR

Commonwealth Edison

9/20173

Two units at each of two sites.
CPs issued 12/31175

Duke Power5/24174

Three units at each of two sites. Also
references CESSAR. Cherokee CPsissued 12/30/77.

Duplicate Plants
Byron 1&2
Braidwood 1&2
Cherokee 1,2&3

Five units at four sites.

SNUPPS
Wolf Creek

Kansas Gas & Electric Co.
Kansas City Power & Light

5/17174

CP issued 5/17/77

Callaway 1&2

Union Electric

6/21174

CPs issued 4/14176

Tyrone 1

Northern States Power

6/21174

CPs issued 12/27177

Sterline

Rochester Gas & Electric

6/21174

CP issued 9/01177

Wisconsin Electric Power

8/09/74

Initially submitted under duplicate plant
option with intent for as many as
six total units at three sites. Utility'schange in plans led to removal fromstandardization program by staff. Review
discontinued because of site problems

Offshore Power Systems

7/05/73

Entire plant design

Jamesport 1&2

Long Island Lighting

9/06/74

Replicates Millstone 3

Marble Hill 1&2

Public Service of Indiana

9/17175

Replicates Byron 1&2

New England 1&2

New England Power & Light

9/09176

Replicates Seabrook 1&2

Palo Verde 4&5

Arizona Public Service

3/31178

Replicates Palo Verde 1,2&3

Haven 1

Wisconsin Electric Power

4/05178

Replicates Koshkonong 1&2

WNP

Koshkonong 1&2

Madison Gas & Electric
Wisconsin Power & Light
Wisconsin Public Service
License to Manufacture
Floating Nuclear
Plant (FNP) 1-8
Replication

98====~======~=====================================

Staff studies (NUREG-0427) have shown that the
NRC standardization program is about at the breakeven point, that is, the staff resources spent on thereview of standardization plants and design approval
applications is about equivalent to the resources thatwould have been used if only custom plants had been
involved. To the extent that utilities reference approved designs in the future, the balance will become more
and more favorable for the standardization program.
On the other hand, should the staff be requested toreview additional PDA's and new applications that do
not reference PDA's, FDA's, or ML's (Manufacturing
Licenses), the use of standardization to reduce the useof staff resources would not be realized.
Staff studies also have revealed that use of the standardization options have not, to date, resulted in a
reduction of schedules. These studies show that thepotential exists for significant schedule reductions only
when there is preapproval of the Nuclear Steam Supply System (NSSS), the Balance of Plant (BOP), andthe site, the three review areas that separately candefine the critical path. Thus, a strong incentive exists
for pursuing site approvals via the Early Site Heview
Program, since approved PDAs now exist for the NSSS
and BOP portions of the plant. Utility-related mattersof the application, such as the quality assurance program or the financial qualifications, generally do notcontrol the overall review schedule.
Program actions completed during fiscal year 1979
included: (a) extending Balance-of-Plant PDAs to a
full3-year term; (b) extending six PDAs to a full5-year
term based upon a completeness review; and (c) issuing a PDA for RESAR-414. Additional reviews andpolicy initiatives were temporarily suspended in April
1979 as a result of the TMI-2 accident. Staff resourceswere re-directed to high priority activities associatedwith the accident-related studies.

ADV ANCED NUCLEAR POWER PLANTS
On April 7, 1977, President Carter issued a statement on Nuclear Power Policy which restated the role
that' nuclear energy was to have in the total energy
prospects of the country. The President's policy would
defer indefinitely the commercial reprocessing andrecycling of plutonium produced in nuclear powerreactors, restructure the U.S. breeder reactor program
to give high priority to alternative designs, and defer
the time when breeder reactors are to be commercialized.
During this reporting period, the NRC has continued its participation in the review and assessment ofa variety of reactor types and fuel cycles being considered by the Department of Energy (DOE) as part of
the Nonprolifereation Alternative Systems Assessment
Program (NASAP); it also continued performing

reviews and providing comments on the studies andassessments being performed under the International
Nuclear Fuel Cycle Evaluation (INFCE) program. In
its reviews and comments, the staff focused on thepotential licensability of these reactor types andassociated fuel cycles, with respect to safety and safeguards concerns and environmental acceptability.
Based on advanced reactor licensing experience andpreliminary safety documents supplied by DOE, thestaff prepared its initial comments on alternative reactors and fuel cycles and forwarded them to DOE in
June 1979. These initial findings are summarized in
the first of a series of reports to Congress published in
October 1979.

Clinch River Breeder Reactor
The status of the staff review of the Clinch River
Breeder Reactor remained inactive throughout theyear and will remain so pending enactment of legislation clarifying the status of the facility.

Fast Flux Test Facility
The Fast Flux Test Facility (FFTF) is a major
LMFBR test faCility which, with a power of 400
megawatts (thermal), will provide an intense field offast neutrons for irradiating fuels and materials in connection with advanced reactor research and development. The facility, which is located about 10 miles
north of Richland, Washington, is owned by the
Department of Energy (DOE) and is not subject tolicensing by the NRC. An NRC staff safety review was
performed, however, under terms of an interagency
agreement with DOE. The staff completed the majorpart of its review effort and, in August 1978, issued its
Safety Evaluation Report (NUREG-0358). A supplement to the SER (NUREG-0358, Supplement No.1)
was issued in May 1979. Sodium filling of one secondary sodium loop took place in July 1978. Fuel loading
was expected in October 1979. Prior to full power
operation, now scheduled for early 1980, a series oftests was to be performed to determine whether
natural circulation is a viable method of removing
decay heat as predicted by analyses.
The Advisory Committee on Reactor Safeguards(ACRS) was extensively involved in the review of
FFTF and meetings addressing that review were heldin July, August, September and November 1978. The
ACRS concluded that the startup and operation of the
FFTF is acceptable, provided that due regard is given
to NRC consequences of certain low probability
accidents, and other specified matters. DOE ispresently evaluating the NRC staff recommendations
regarding containment adequacy for low probability
accidents.

99

Gas-Cooled Reactors
As a consequence of the withdrawal of the General
Atomic Company from the commercial nuclear power
market in late 1975, regulatory activities related togas-coole'd reactors have been confined primarily to
the Fort St. V rain reactor. Limited reviews ofadvanced high-temperature gas-cooled reactors and of
a gas-cooled fast, breeder reactor have been undertaken in conjunction with the NRC's participation in
the NASAP study.

Fort St. Vrain. Fort St. Vrain, a 330-MWe hightemperature gas-cooled reactor (HTGR), was designedby the General Atomic Company and is operated by
the Public Service Company of Colorado near
Platteville, Colorado. Transfer of ownership to PublicService was made in June 1979. Power level isrestricted to 70 percent of initially rated power pending resolution of the fluctuation problem described on
page 40 of the 1978 NRC Annual Report.
Advanced High-Temperature Gas-Cooled Reactors.
In early 1978, a group of utilities formed an organization, Gas Cooled Reactor Associates (GCRA), for the
purpose of developing a commercially viable advanced
HTGR. GCRA manages the DOE funds supporting
the project and is responsible for carrying out initial
phases of the licensing review. In early 1979, a decision was made to terminate work on a standardized
900 MWe steam cycle plant in favor of' working
toward the demonstration of the gas turbine cycle inthe mid-1990·s. NRC review of this concept is being
performed under NASAP auspices.
Gas-Cooled Fast Breeder Reactor. In late 1976, anorganization of utilities, Helium Breeder Associates(HBA) , was formed to work with both the General
Atomic Company and DOE (then the Energy
Research and Development Administration) toward
the development and demonstration of the Gas-Cooled
Fast Breeder Reactor (GCFR). The GCFR demonstration unit would produce 330 MWe. Both DOE andHBA have accented General Atomic's revised reactor
design that would permit emergency core cooling bymeans of natural convection. This concept is now
being reviewed under NASAP auspices.

Floating Nuclear Power Plants
Floating nuclear power plants (FNPs) are electricalgenerating stations of a standardized design which
would be constructed at a shipyard facility using
assembly line techniques. The proposed FNPs would
utilize a conventional pressurized light water reactor
system design mounted on floating platforms, similar
to the hull of a barge, and can be sited at offshore ornearshore sites in the ocean or in estuaries and rivers.
Offshore Power Systems (OPS) , a subsidiary ofWestinghouse Electric Corporation, filed an application with the NRC in 1973 for a license to manufacture

up to eight identical floating nuclear power plants at
Blount Island near Jacksonville, Fla.
An NRC staff Safety Evaluation Report(NUREG-75/100) was issued in September 1975; Supplement No. 1 (NUREG-0054) was issued in March
1976 and Supplement No.2 in October 1976. It was
anticipated that Supplement No. 3 will be issued inearly 1980.
The staff has also prepared a three-part Final Environmental Impact Statement (FES) to assess thepotential impacts from the construction, siting andoperation of FNPs. Part I, issued in October 1975,
relates to the construction and nonnuclear testing of
the FNPs at the manufacturing site in Florida. Part I
concluded that foreseeable adverse impacts frommanufacturing the FNPs would be acceptable in consideration of the benefits expected from the plants andtherefore recommended that a manufacturing license
be issued, subject to certain license conditions. Part II,
issued in September 1976, relates to' the potential impacts associated with siting, constructing, andoperating FNPs at generalized unspecified locations
offshore in the Atlantic Ocean, Gulf of Mexico and incertain rivers and estuaries. At the request of theCouncil of Environmental Quality, the NRC issued an
Addendum to Part II of the Final Environmental
Statement, in June 1978, which elaborated upon thedata and analyses presented in Part II with respect tothe estuarine and riverine siting of FNPs. The staff
concluded there was reasonable assurance that eight
FNPs could be sited, constructed and operated withacceptable environmental impact at offshore sites
along the Atlantic Ocean and Gulf of Mexico and at
carefully selected shoreline locations, including
estuarine waters. The U.S. Environmental Protection
Agency, however, believes that it will be difficult tofind environmentally acceptable sites in any of theestuarine or barrier island areas along the East andGulf Coasts.
Part III of the FES, issued in December 1978, compared the total risk to the environment from accidental releases of radjoactivity for both floating and landbased nuclear power plants. A wide spectrum of accidents were considered including, for the first time,
low-probability, core-melt accidents (Class 9) in the liquid pathway. Part III also included an overall costbenefit analysis for all elements of the environmental
statement and concluded that a manufacturing license
should be issued subject to several license conditions,
including a specific license condition to mitigate thepotential environmental impacts from a core-melt accident at an FNP. This involves the use of a material
such as magnesium oxide beneath the reactor vessel inorder to retard the penetration by the melting corethrough the bottom of the FNP hull.
Part III also listed NRC generalized requirements
for compliance by a utility/operator of an FNP when

100

an application is made to the NRC for locating an FNP
at a specific site. These include modification of FNP
sites in estuaries, rivers or near barrier islands so as tolimit the release of radioactive materials into the surrounding water body following the unlikely event of acore-melt accident. A principal reference used in the
preparation of the FES, Part III was the Liquid
Pathway Generic Study (NUREG-0440), issued in
February 1978.
Public hearings on safety and environmental issues
were started in March 1975 and continued through1979. Offshore Power Systems appealed the staffs
precedent-setting decision to include Class 9 accidents
in the comparative analyses of the FES. In December
1978, the Commissioners agreed to review whether
Class 9 accidents were a proper subject for treatmentin the environmental impact statement. On September
14, 1979, the Commission issued a Memorandum andOrder in the Matter oj Offshore Power Systems whichstated the Commission's position that the staffs
analysis of the Class 9 accident question is properly included in the environmental impact statement in thisproceeding in order to meet NRC's NEPA responsibilities. Both the applicant and the staff have submitted partial proposed findings of fact to the licensing
board and all safety and environmental contentions
have been addressed during the hearing process. Additional hearings were held in late 1979 in order todiscuss the licensing board's questions regarding thestaff's Class 9 analysis.
The first application for a permit to construct andoperate an offshore floating nuclear power station was
filed in 1973 by the Public Service Electric and Gas
Cmnpany (PSE&G) of New Jersey. The proposed
Atlantic Generating Station (AGS) consisted of two
floating units (1150 MWe each) located approximately
three miles off the coast of New Jersey and about 11
miles northeast of Atlantic City. In December 1978,
PSE&G cancelled its contract with OPS, citing amongits reasons the lower than anticipated electricity
growth rate in its generating area. The application has
been withdrawn and the licensing proceeding dismissed.

PROTECTING THE ENVIRONMENT
Health Effects of the
Coal and Nuclear Fuel Cycles
As noted in the 1978 Annual Report, the NRC is actively involved in developing estimates of potential effects of the coal and nuclear fuel cycles to aid in theanalysis of alternative energy sources for generating
electricity. Final revision of the draft report, "Health
Effects Attributable to the Coal and Nuclear Fuel
Cycles" (NUREG-0332), is being held in abeyancepending release of the latest National Academy of
Sciences Report of the Committee on Nuclear andAlternative Energy Sources (CONAES). A contracted
study with the Argonne National Laboratory on healtheffects models for the nuclear and coal fuel cycle alternatives is nearing completion.
In November 1979, the staff issued a report
prepared under contract by Teknekron, Inc., entitled,"Activities, Effects and Impacts of the Coal Fuel Cycle" (NUREG-CR-I060). The report provides a current data base related to the health, ecological,
economic and social impacts of the coal fuel cycle. The
report considers the impacts resulting from the various
phases of the coal fuel cycle: resource extraction, coalcleaning, transportation, storage, power production
and waste disposal.

Assessment of Radiological Consequences
Of Radionuclide Releases
By means of Federal Register notice of January 13,
1977 (42 FR 2858), the Environmental Protection
Agency officially issued 40 CFR Part 190, Environmental Radiation Protection Standards jor
Nuclear Power Operations. The standards require that
operations covered by the subpart B, Environmental
Standards jor the Uranium Fuel Cycle, shall have noplanned discharges that will result in an annual dose
equivalent to any member of the public that will be inexcess of 25 millirems to the whole body. Other requirements involve organ doses and releases of specificradioisotopes. The standards are effective as ofDecember 1, 1979, except that for doses arising fromoperations associated with the milling of uranium ore,
the effective date is December 1, 1980. For releases ofiodine-129 and krypton-85, the standard will be effective January 1, 1983. The standards may be exceeded
during a given year of operation only if the regulatory
agency has granted an exemption based on a determination that a temporary and unusual operating condition exists and that continued operation is in the public
interest.
The NRC has been developing provisions to be incorporated in license conditions requiring that NRC
licensees meet the conditions of Part 190. Most nuclear
power plants that meet the requirements on radioactive effluents promulgated by Appendix I to 10 CFRPart 50 have been shown generically to meet Part 190.
To assure full compliance, the model Radiological Effluent Technical Specifications (RETS), contained in
NUREGs-0482 and -0473, have been modified to include Part 190 requirements as a limiting condition for
operation. Staff documents (NUREGs) further
describing acceptable methods for demonstratingcompliance with Part 190 are in progress. The Officeof Nuclear Material Safety and Safeguards is develop-

====================================================101
ing license conditions for the uranium fuel cycle
facilities under its cognizance. The Office of Standards
Development is preparing modifications for Title 10
regulations and for regulatory guidance documentsthat will further identify the requirements that Part
190 places on NRC licenses.

Control of Effluents
Standard Technical Specifications. As a result of thestafFs continuing review and discussions with theAtomic Industrial Forum and other parties of interest,
substantive revisions were made to the NRC draft
reports on "Radiological Effluent Technical Specifications" (NUREG-0472 for PWRs, and NUREG-0473 for
BWRs). The revised reports still incorporate the fundamental requirements and concepts contained in theoriginal, but equations for dose calculations, setpoint
determination, and meteorological dispersion factors
have been eliminated. These equations, among otheritems, are now required to be included in an Offsite
Dose Calculation Manual (ODCM) that is to be provided by each licensee to NRC for review and approval
along with the proposed Technical Specifications.
Regional seminars were held in late 1978 to provide
guidance in the preparation of the Technical
Specifications. All affected utilities were invited tosend representatives to these seminars. Licensees are in
the process of submitting their proposed Technical
Specifications and review of these by the staff was inprogress at the close of the report period. Licensee submissions and NRC reviews should be completed by
mid-1980.

would be kept at levels as low as possible under the circumstances and to remain within established NRC
effluent standards (see Chapter 2).
During the long-term recovery period for the TMI
plant, the TMI-2 Support Task Force of the NRC willcontinue to review all matters related to maintenanceof the reactor in a safe shutdown condition, decontamination of equipment and buildings, installation of
new equipment and systems, the processing of liquids
contaminated from the accident for removal of radioactivity, the storage and shipment of radioactive
wastes, and releases of low levels of radioactivity to theenvironment. Safety evaluations and environmental
assessments for the more significant recovery operations are being prepared. These activities are being
coordinated with local, State, and other Federal
officials.

Site-Related Problems

In-Plant Measurements Program. In order to assurethat the best available data is employed in improving
the calculational models used to appraise conformanceof licensees with Appendix I to 10 CFR Part 50, the
NRC contracted with Idaho National Engineering
Laboratory to perform in-plant measurements onpressurized water reactors (PWRs). The measurements
will provide a data base for radidoisotope inventories
in plant systems, radioactive waste management
system performance, and source term for both liquidand gaseous systems. As of the end of the reportperiod, measurements had been completed at four
plants (Zion, Fort Calhoun, Turkey Point, andRancho Seco).

Rejection of Greene County Site Due to Esthetic Impacts. It was during 1979 that the NRC for the firsttime rejected a proposed nuclear site primarily
because of adverse socioeconomic impacts. The staff
concluded that the proposed Greene County plant at
the Cementon, New York site would result in unacceptable esthetic impacts on certain local, regional, andnational historic, scenic, and cultural resources. The
major reason for that judgment was the visual intrusion of plant facilities-primarily the natural-draft
cooling tower and its plume-into the central view
from Olana, the home of 19th Century painterFrederick Church, which had been designated as a
National Historic Landmark. Other visually sensitive
areas that could be adversely affected were also identified. The staff analyzed the esthetic impacts fromalternative cooling systems but determined that eventhe least visually obtrusive alternative would still belikely to have undesirable effects. The staff also identified severe (although generally mitigable) socioeconomic impacts arising from the potential loss of a
local industry because of the facility's land needs andfrom the need to substantially change the local
transportation network to serve the facility. The staff
concluded that there are several alternative sites in
New York State that, on the basis of environmental
considerations, are obviously superior to the proposed
site.

Three Mile Island Accident Response. The performances of the effluent treatment systems following theonset of the TMI accident were evaluated and theamounts of released radioactivity have been assessed.
Recommendations were made for operating procedures involving the use of existing equipment and theinstallation of new equipment needed to assure that
releases of radioactivity during this emergency period

Mass Mortality of Biota. NRC continues to monitor
potential environmental problems arising fromoperating nuclear power plants. In the summers of1978 and 1979, a sizeable number of weakfish (alsoknown as sea trout) were drawn onto the intake
screens at the Salem Nuclear Station. At the Oyster
Creek Station in August 1979, unusually high temperatures resulted in the apparent loss of a small but still

102

significant number of several species of fish. The staff
is working closely with EPA in reviewing the facts
associated with these occurrences to determine
whether corrective action should be taken.
Water Quality Monitoring Requirements. In order
to ascertain the environmental consequences of power
plant licensing, NRC is placing increasing reliance on
EPA's permit system, a result of the National Pollution
Discharge Elimination System (NPDES). A major stepto avoid the confusion and inequity resulting from
regulation of the aquatic environment by two Federal
agencies was taken with the closely coordinatedreview of TVA's Yellow Creek Nuclear Station Construction Permit Application. As a consequence of the
Yellow Creek Proceeding, which suggested that thisapproach was not only desirable but legally necessary,
the NRC staff is striving to obtain EPA or State agency
resolution of questions pertaining to water quality that
may arise during NRC's environmental review.
Implementation of Executive Order 11988 on
Floodplains. By an Executive Order issued in May
1977, President Carter called upon Federal agencies toconsider any action they undertake affecting thenation's floodplains as an opportunity to reduce theimpact of floods on human safety, health and welfare,and to restore and preserve the natural and beneficial
values served by floodplains. The NRC staff developed
procedures for reviewing reactor sites in a mannerconsistent with the intent of the Executive Order andpublished these in the Federal Register on October 6,
1978. In addition, licensing procedures and Environmental Standard Review Plans were revised to address
floodplain issues more explicitly. During 1979 the staff
undertook the evaluation of several reactor sites with aview to improving floodplain management. Most sites
for nuclear power plants require placement of some
type of facilities in floodplains, such as auxiliarybuildings, pipelines, and roadways associated with intake and discharge structures. Usually they are smallin size, relative to the floodplain cross-sectional area,and do not interfere significantly with its floodhandling capability. If significant impacts are identified, it is generally required that structures be
relocated or redesigned, or that other measures betaken to preserve the floodplain function.
The Executive Order on Floodplain Management
also requires that floodplain considerations be addressed in NRC environmental impact statements and that
guidance be afforded to applicants so that they canevaluate the effects of their proposals on floodplainsprior to submitting their applications. The NRC's Environmental Standard Review Plam, published in May
1979 as NUREG-0555, satisfy these requirements forenvironmental concerns. Specific instructions are
given for the NRC staff analysis of potential floodplain

impacts and the discussion of these impacts in theCommission's Environmental Statements. A separate
portion of each environmental standard review plan,
describing data and information sources needed to
conduct the environmental review, may be used byapplicants as a guide for the treatment of floodplain
concerns in their environmental reports.
Evaluation of Breakwaters for Coastal Nuclear
Plants. Breakwaters made up of massive rocks or concrete armor units are often used to protect nuclear
power plants from the effects of damaging waves. In
1979, the staff studied two special cases involving
breakwaters and related to both safety and environmental issues. The breakwater at Pilgrim Nuclear
Generating Station near Plymouth, Massachusetts had
been damaged by the same storm which produced
record snowfalls in the Boston area in February 1978.
The breakwater was damaged again during the winterand early spring of 1979. The staff evaluated the probable causes of the damage, inspected the repair of thestructure, and is currently evaluating the possible
redesign of the breakwater to assure that the plantsafety systems are not compromised.
Breakwater designs for floating nuclear plants werealso studied with a view to assuring safety of the plant
and to minimizing environmental impacts to oceans
and estuaries. For estuarine siting, the staff identified
a special problem which might occur in the event of an
accidental release of radiation through an open
breakwater. In estuaries, the mixing and dispersion ofradioactive liquid from the accident would be slow
and would produce long-term radioactivity levels withsevere effects on the biota. The staff has recommended
that estuary sites be such that the consequences of
postulated accidents will not be worse than they would
be at comparable land-based sites.
Modification of Environmental Technical Specifications for Operating Reactors. In accordance with provisions of the National Environmental Policy Act(NEPA), NRC has included in operating licenses a requirement for environmental monitoring programs. A
number of fully licensed operating stations have compiled five years or more of data, and the staff has
found, in reviewing the records of several such stations, that actual environmental impacts are generally
within the range of expectations set out in theEnvironmental Impact Statement prepared prior tolicensing. Review of other stations' results is expected
to provide significant feedback to the license application review process.
Adequacy of Analysis of Alternative Sites. An important means of protecting the environment against
undue adverse impacts is an appropriate screening of
the area, as part of the site selection and evaluation
process. Sufficient effort must be made by the utility
applicant and the NRC review team to identify those====================================================103
alternative sites which are among the best which
reasonably could have been found and to identify andproperly assess the major potential environmental impacts. Such an effort is esseQtial to a staff determination that no alternative site is obviously superior to thesite proposed by the applicant. Without thatreasonable assurance on the part of the staff, the site is
to be rejected. In the Limited Early Site Review for
the proposed Perryman nuclear plant site about 20
miles northeast of Baltimore, the NRC staff determined that the applicant's site selection and evaluation
methods were inadequate and also concluded, on thebasis of its own reconnaissance-level investigations,that there were obviously superior alternative sites.
The staff conclusion that there are sites which are obViously superior to the Perryman site was based on
considerations of population density, risks posed by
the proximity of potentially hazardous activities andthe overall project costs.
The review of the Perryman and other sites considered by the applicant failed to identify any environmental considerations that would suggest that
Perryman offers Significant advantages over alternative sites. The applicant did provide information to
support his view that there were economic advantages
in locating at Perryman, primarily resulting fromlower transmission costs. In the staffs view, this cost
avant age would be more than offset by the special
design provisions which the staff expects will be required to protect against nearby external hazards.
Among other reasons, the staff concluded that the applicant's alternative site analyses were inadequate
because important siting parameters were omitted. In
particular, no demographic or safety-relatedcharacteristics were used in the comparisons amongsites. The process used to select, and the scoring
scheme used to compare, the relative merits of the candidate sites have a number of deficiencies which
render the results unreliable. Together, these factors
cast doubt as to whether the submitted candidate sites
represent the realistic siting resources available to theapplicant. Following this review, the staff has madesubstantial progress during 1979 in formulating review
policies to improve the adequacy of analysis of alternative sites.
Progress in the staffs reevaluation of considerations
which are important to a determination of the adequacy of alternative site analyses was also stimulated
by licensing actions involving Seabrook Units 1 and 2
and Pilgrim 2. During 1978, the NRC staff wasordered by the Atomic Safety and Licensing Appeal
Board (AS LAB) to reexamine the question of alternative sites to the Seabrook Station located in Seabrook,
N. H. The study was conducted under the assumptionthat the Environmental Protection Agency mightorder the construction of cooling towers for that station rather than permit once-through cooling; a valid

alternative site analysis had already been done for thelatter design and it had been determined that the
Seabrook site was the environmentally superior site if
cooling towers need not be used. The staff st':ldied 22
candidate sites located throughout New England. Using a coarse screening process, the staff reduced thisnumber of alternative sites to eight, and these werethen compared in considerable detail to Seabrook. The
analysis involved the contributions of eight different
environmental disciplines. In addition, assistance wasobtained from Argonne National Laboratory and Oak
Ridge National Laboratory in computer modeling ofcooling towers. Assistance from the Federal Energy
Regulatory Commission was obtained for the analysisof transmission stability and reliability at each site, incomparison with Seabrook. After extensive comparative analyses, the staff concluded that five sites
were environmentally equivalent to Seabrook and one
had minor environmental advantages, while the remaining sites were environmentally disadvantageous.
None of the alternative sites was found to be "obviously superior" to Seabrook under an assumption that
Seabrook Station would be required to have cooling
towers. The staff presented its testimony in a three-day
hearing before the ASLAB in January 1979. Subsequently, the EPA reached a final decision that oncethrough cooling at the Seabrook Station was enVironmentally acceptable and the issue of alternative
sites with cooling towers at Seabrook became moot.
As for Pilgrim 2, on December 1, 1977, the Atomic
Safety and Licensing Board issued a partial initialdecision regarding only the alternative site analysis
section of the environmental review. The board's decision denied the Limited Work Authorization requestedby the Boston Edison Company, citing as its reason theinadequacies of the NRC staffs review of alternative
sites. This decision necessitated a reevaluation ofalternative sites by the staff which, in turn, led to a request of the applicant to provide supplemental information on alternative siting. The staff conducted adetailed review of 13 sites located in Massachusetts,
New Hampshire, and Connecticut. Each alternativesite was evaluated against the Pilgrim site with respect
to prospective impacts in the area of aquatic biology
and water quality; terrestrial ecology and land use;
demography; adjacency to industrial, transportationand military facilities; hydrology; socioeconomics;
project economics; geology, seismology andgeotechnical engineering; and meteorology. The staff
analysis was presented in a Final Supplement to the
Final Environmental Statement, issued in May 1979.
As a result of this more detailed analysis, the staff concluded that none of the alternative sites was obviously
superior to the proposed Pilgrim site and therefore
recommended acceptance of the proposed site for thesecond unit.

104================================================
Reasonably close proximity to an adequate watersupply and the likely extent of water-related environmental impacts are generally regarded as among
the more important factors in the identification of asuperior site for large baseload electrical generating
facilities. The cost and effectiveness of controls for
mitigating such impacts is also examined. In both the
Seabrook and Pilgrim 2 reviews, a number of waterrelated aspects were given detailed analyses for the
proposed and alternative sites: local drainage considerations; erosion control; flood protection; pipeline
location for coolant water; location of intake anddischarge structures; water supply availability; andpossible contamination of water supplies. In bothcases, reviews of these considerations revealed thatmost were not critical to a demonstration of obvious
superiority among the final group of candidate sites.
This was so because site-screening criteria had alreadyeliminated the most objectionable sites in terms ofadverse water-related impacts, and because a numberof the adverse water-related impacts for the final
groups of candidate sites analyzed in detail could
readily be mitigated at reasonable cost.
The major exception to this conclusion was thatwater availability remained a key siting issue. In particular, for inland sites on rivers or streams whereflows may be seasonably very low, careful consideration will be required to assure that an adequate anddependable supply can be provided to meet thecoolant water needs of the generating plant and ofother users in the region.

NRC has asked for formal consultation with the National Marine
Fisheries Service conceming the possible impact of nuclear power
plants on populations of shortnose sturgeon in the Delaware River.
The sturgeon shown here was recovered from the water intake

INTERAGENCY COORDINATION
Compliance with Regulations
Of Other Agencies
Cooperation with EPA and DOE on Occupational
Radiation Dose Limits. NRC staff members are participating in an Environmental Protection Agency(EP A) Interagency Committee on Federal Guidance
for Occupational 'Exposures to Ionizing Radiation.
The objective is to assist the EPA in developing
guidance on occupational dose limits, responding inpart to the recommendations in Publication No. 26 of
the International Commission on Radiological Protection. In a related development, the NRC staff is working with EPA and the Department of Energy (DOE) in
response to petitions by a private citizen and the
Natural Resources Defense Council. Among othermatters, these petitions request significant reductionsin the occupational dose limits. EPA and N~C are considering joint hearings on these subjects to provide an
opportunity for interested parties and members of the
public to participate and to make their views known.
- Floodplain Management. In accordance with theprovisions of Executive Order 11988. "Floodplain
Management," NRC has consulted with the Federal
Interagency Panel on Floodplain management concerning procedures for floodplain management associatedwith power plant licensing applications (see discussion
earlier in this chapter).
Endangered Species Act (ESA). This. Act wasamended in 1978 to require Federal agencies to consult

system of the Salem Nuclear Generating Station, Unit 1, in New
Jersey. Recent surveys indicated that the species, once threatened,
may be making a comeback.

====================================================105
with the Fish and Wildlife Service or the National
Marine Fisheries Service (NMFS) to determine thepossible existence of any endangered species in thelocality of proposed projects prior to their initiation. If
listed species are in the project location, biologicalanalysis and further consultation are required.
However, if none is found, no further interagency consultation is needed.
In 1978 two shortnose sturgeon had been found on
the intake structure at Salem 1, and, in 1979, the NRC
requested formal consultation with the NMFS to
determine whether operation of the plant jeopardized
the continued existence of the species in the Delaware
River. Increasing numbers of reports of short nose
sturgeon in recent years indicate that the species may
be making a comeback, not only in the Delaware, but
in other Atlantic coast riverine systems as well. Formal
consultation has also been requested on the shortnose
sturgeon at Salem 2 and Hope Creek 1 and 2.
Shortnose sturgeon are also found at the Indian
Point (New York) and Hatch (Georgia) plants. At Indian Point, the status of the short nose sturgeon isunder review by the NMFS. At Hatch, the NRC has initiated actions required by Section 7 of the ESA.
For other operating facilities, consultation under
the ESA will be initiated when the staff becomesaware of the existence of threatened or endangered
species in the plant vicinity.
Fish and Wildlife Coordination Act. During 1979,
the staff reviewed proposed regulations of the Fish andWildlife Service for implementation of the Fish andWildlife Coordination Act. The regulations, if
adopted, would require Federal agencies to mitigate
significant impacts on fish and wildlife in all waterrelated projects and to compensate for significant
wildlife losses or enhance wildlife values for all
Federally licensed projects. Consultation between
NRC and the Fish and Wildlife Service would be required by the regulations in all reactor licensing.
Coastal Zone Management Act. Under this legislation the siting of a nuclear plant in a region bordering
the seacoast or the Great Lakes must be consistent
with the Coastal Zone Management Plan developed by. each State, commonwealth or territory in which thesite is located. The State or commonwealth (e.g., Puerto Rico) makes this determination. As of the end of
fiscal year 1979, over half of the 30 affected States
have completed or made substantial progress indeveloping management plans for their coastal zones.
Clean Water Act. The environmental review of
NRC licensing actions involves extensive coordination
with other Federal and State agencies concerning provisions under the Clean Water Act. A principal area of
coordination is with the EPA or delegated State agencies under the National Pollutant Discharge Elimination System Permit Program. NRC's relationship to

EPA was modified in 1978 by the Yellow Creek decision of an NRC licensing Board which required that
NRC coordinate its nuclear plant water quality
monitoring needs through the EPA or "permitting"State, rather than impose water monitoring requirements of its own directly on applicants. EPA isrevising its Effluent Limitation Guidelines in conjunction with quality standards, which will regulate theconcentrations of non-radiological contaminants innuclear power plant effluents.
Toxic Substances Control Act. Guidelines developed
under the Toxic Substances Control Act may result in
the imposition of limitations by EP A on certain
substances found in power plant cooling water
discharges. On the list of substances currently being
evaluated by EPA are several which are produced bychlorination of natural surface waters used for power
plant cooling.

Interagency Topical Studies
Interagency Committee on Ocean Pollution
Research, Development, and Monitoring. This committee was created by Public Law 95-273' to prepare
for fiscal years 1979-1983 a Federal Plan for Ocean
Pollution Research, Development and Monitoring.
The final draft of the plan was completed in August
1979 and submitted to the Executive Office of the
President by NOAA, the study coordinator. The study
identified current Federal ocean pollution research activities and established a prioritized program forresearch and development. The study also initiatedplanning for coordination of future ocean pollution
research and for dissemination of the information
resulting from the research and monitoring programs.
The NRC funded approximately $1.2 million of ocean
pollution research in fiscal year 1979 to support its. licensing actions. In addition, considerable monitoring of the ocean environment has been performed byapplicants for nuclear power plant construction permits and licensees of operating nuclear plants. The
next version of the Federal Plan on Ocean Pollution
Research will include work being done by private andState organizations as well.
Interagency Committee on Environmental Monitoring. At the direction of the President, the Council on
Environmental Quality established the Interagency
Task Force on Environmental Data and Monitoring.
Its function is to review environmental monitoringand data programs and to recommend improvementsthat would make these programs more effective.
Specific activities of the Task Force in which NRC participates consist of assessing the manner in which thevarious Federal agencies accumulate and disseminate
water and air data, developing a catalog of agency
monitoring, and otherwise coordinating the use of this106

information among other Federal agencies. Recommendations were made in a report to CEQ that is to be
forwarded to the President. The primary purpose of
these recommendations is to minimize overlapping
responsibilities among the Federal agencies and to provide for communication and coordination of datacollection efforts related to water and atmosphericproperties and constituents.
TVA/EPRI Workshop on Waste Heat Utilization.
During the report period, the NRC staff participatedin this workshop sponsored by the Tennessee Valley
Authority and the Electric Power Research Instituteby chairing the session on nuclear and public healthaspects of waste heat utilization, and by presenting ananalysis of the specific factors that would be of concern to NRC in the use of effluents from nuclear power
plants. The results of this workshop will be applicableduring the Watts Bar licensing process because of theapplicant's plan to provide for a large waste heatutilization facility associated with the power plant.
Water Resources Council. During the report period,
the staff participated in activities of the Hydrology
Committee of the Water Resources Council. The activities of the committee centered on coordinating thevarious participating agencies working to assess thestate-of-the-art in various subject areas and to recommend standardization. The staff participated inseveral activities in this area, including an assessmentof the state-of-the-art in hurricanes surge modeling(Le., induced flooding), reassessment of groundwater
study requirements, assessment of the state-of-the-art
in low flow considerations, and in identifying standardization of flood frequency determinations forungaged water sheds.
Interagency Advisory Committee on Water Data.
The staff participated on the Interagency Advisory
Committee on Water Data chaired by the U. S.
Geological Survey. During fiscal year 1979, the staff
identified NRC's water data uses and needs andcooperated with other Federal agencies under the provisions of OMB Circular 67 on water data.
Interagency Committee on Dam Safety. In April
1977, President Carter requested a review of procedures and criteria issued by Federal agencies involved in the design, construction, operation, and regulation of dams, and the preparation of guidelines for
management procedures to ensure dam safety. The
guidelines, published in June 1979, are based on an intensive review of agency practices conducted by three
department groups: the departments and agencies
themselves; an ad hoc interagency committee of theFederal Coordinating Council for Science, Engineering and Technology (FCCSET); and an Independent
Review Panel of recognized experts from the academic
and private sectors. NRC was requested to join in the

FCCSET review and guideline development activity.
The guidelines-which address organization, management, and the management of technical activities, including site investigation and design, construction,and operation and maintenance-await Presidential
direction for implementation.
Interagency Committee on Seismic Safety in Construction. NRC staff participated principally on a subcommittee of the Interagency Committee dealing withthe problems of tsunami (seismic sea wave) protection.
This subcommittee's purpose is to identify methods
and criteria for assessing tsunami and otherseismically-induced flood wave threats for the protection of Federal facilities. State-of-the-art in assessing
tsunami threats and design criteria were drafted infiscal year 1979.
The National Weather Service Study on Emergency
Response. Subsequent to the accident at Three Mile
Island, the staff coordinated with its consultants, theNational Weather Service of the National Oceanic andAtmospheric Administration (NOAA), and DOE in an
attempt to determine the actions taken by NOAA and
DOE during the TMI accident. The NRC staff
response was independent of both NOAA and DOE,
and it was only subsequent to the accident that thestaff learned of the extent of both NOAA's and DOE's
participation. These facts exposed the need to coordinate future agency responses to any accidents involving core melt with significant offsite radiation releases(Class 9). Moreover, the staff has requested NOAA's
participation in a study of portable meteorological instrumentation and of an assessment of meteorologicalmodels for use during any such future accidents. The
purpose of the instrumentation and models would beto supplement on-site and regional meteorological
data and models to provide prompt and accurate
estimates of the location and concentrations ofradioactive releases.
Interagency Committee on Electric Field Effects
from High Voltage Transmission Lines. NRC participates as a permanent member on the Interagency
Advisory Committee on Electric Field Effects and aidsin the identification, review, and coordination ofFederal research programs investigating potential problems associated with the operation of electricaltransmission systems. The Committee's primary emphasis has been on the study of potential short andlong term health effects from the operation of highvoltage transmission systems. These effects areroutinely considered by NRC staff in its environmental
reviews.
National Ecological Assessment Workshops. NRC
participated in a workshop sponsored by the U.S.
Department of Energy for the purpose of ideptifying

=================================================107
the ecological concerns, methods and problems of performing non-site-specific ecological assessments ofenergy development. The workshop which was attended by prominent ecologists and government planners deliberated for three days on the possible
ecological and human environmental impacts ofenergy development alternatives. Concerns about additions of carbon dioxide, sulfur dioxide, and nitrogenoxides to the atmosphere and widespread acid rainfall
were among the serious problems identified anddiscussed. Although nuclear plants produce none ofthese atmospheric pollutants, the impacts ofgenerating electricity from fossil or other organic fuels
include such pollutants and fall within the scope of
NRC reviews of the comparative impacts of alternative sources of energy for baseload generation ofelectricity.
Third Conference on Water Chlorination-Environmental Impact and Health Effects. The NRC has
joined the Department of Energy, the Environmental
Protection Agency, the National Cancer Institute, The
Tennessee Valley Authority, and Oak Ridge National
Laboratory in sponsoring the Third Conference on
Water Chlorination. The intentional use of toxic
substances to control biological growth within power
plant condenser cooling systems is carefully examined
during NRC's review of a power plant license application and the preparation of an environmental impact
statement. The series of conferences on chlorination
provide current information necessary for the better
understanding of environmental effects and of alternative "bio-fouling" control practices.

COOPERATION WITH STATESState Participation in NRC's
Environmental Impact Statements
During the past year, NRC and the State of New
York took steps which led to the signing of an agreement that allows State participation in the preparationof NRC's environmental impact statements. By termsof the agreement the technical staff of the New YorkState Public Service Commission will write specificsections of the Draft and Final Environmental
Statements. The sections involved are mainly concerned with environmental description and impact, alternatives to the proposed action, and the need for thefacility. In the past, these sections have been preparedthrough environmental assistance agreements with the
Department of Energy's National Laboratories. In addition to the staff of the Public Service Commission,
the technical staff of the New York State Departmentof Environmental Conservation has agreed to provide
technical support to the Public Service Commission in
the form of review and comments on the write-ups forthe environmental statements. State participation ofthis kind in the preparation of NRC's Environmental
Impact Statements will require close coordination
with the relevant review functions of the NRC
regulatory staff.

IMPROVING THE LICENSING PROCESS
Generic Rulemaking to Improve Licensing
In June 1977, an NRC study group seeking to identify ways to improve the effectiveness of NRC Nuclear
power plant licensing procedures recommended,among other things, that rulemaking should be considered for the generic resolution of certain major
issues that are presently litigated in individuallicensing proceedings. The study groups recommendations
are presented in a report, "Nuclear Power Plant Licensing: Opportunities for Improvemenf' (NUREG0292). In response to a Commission directive, the staff
prepared an interim statement of general policy andplans for rulemaking, which the Commission approved for publication in the Federal Register (December
14, 1978). This interim policy statement fully supports
Executive Order 12044 of March 23, 1978, requesting
improvement of existing Federal government regulations, so as to make them as simple and clear as possible and to avoid imposing unnecessary burdens on theeconomy, on individuals, on public and privateorganizations, or on State and local governments. Theinterim policy statement and supporting discussions
are presented in an NRC report, "Preliminary Statement on General Policy for Rulemaking to Improve
Nuclear Power Plant Licensing" (NUREG-0499).
Ten candidate issues were identified by the staff forgeneric rulemaking: (1) future availability and price ofuranium, (2) alternative energy sources to the nuclear
option, (3) need for adding baseload generating
capacity, (4) methodological and information requirements in the analysis of alternative sites (5)criteria for the assessment of nuclear plant impacts
and mitigative measures; (6) generic procedural
criteria to define more concretely NRC responsibility
in assessments and decisions regarding certain waterrelated impacts in relation to the statutory authorities
of EPA and permitting States, (7) NEPA decision
criteria for OL reviews, (8) occupational radiation exposure control, (9) generic radiological impact for normal light water reactor radionuclide releases, and (10)threshold limits for generic disposition of cooling
tower effects. Criteria developed by the Steering Committee on Reactor Licensing Rulemaking to aid in
identifying suitable candidate issues for rulemaking
include the following: the issues must be generic; there
must be a likelihood that a useful, definitive rule canbe formulated and there must be a likelihood that astable rule can be formulated. Value-impact criteria
for appraising the desirability of, and the priorities

108

associated with, specific proposals for generic rules include:• Achievement of more effective public input andimproved public understanding of NRC's
analytical procedures and decision criteria intreating potential environmental and safety issues
in the licensing process for nuclear power plants.
• Improvement of the stability and predictability of
the licensing process, including the provision oforderly and clear procedures for State-Federal
cooperation in treating generic licensing issues.
• Accomplishment of an overall savings of manpower and financial resources of the NRC, thepublic, the utility industry, and other local,
State, and Federal agencies involved in the
nuclear licensing process.
• The short-term increase in dollar costs of thevarious participants in the rulemaking action, including contractual support.
• The additional impacts (Le., opportunity costs) of
diverting manpower and other resources to therulemaking process and away from other productive uses for a temporary period.
Public comment was invited on the merits of thecandidate issues for generic rulemaking and related
decisions criteria, and additional suggestions for candidate subjects for generic rulemaking were solicited.
Fifty-eight comments were received but, except for
Issue No.4 on alternative siting methodology, further
NRC activity on generic rulemaking was temporarilysuspended because of the diversion of staff effort tostudies related to the Three Mile Island nuclear accident and related remedial measures.

Evaluation of Alternative Sites
As noted above, one of the ten issues identified forpossible generic rulemaking was that of alternative sitemethodology and information requirements. In orderto refine and clarify this issue, the staff, on December
14, 1978, issued for comment a report entitled,"General Considerations and Issues of Significance on
the Evaluation of Alternative Sites for Nuclear
Generating Stations under NEPA" (NUREG-0499,
Supplement 1). In addition to receiving public comments on the report, the staff conducted a three-day
public workshop in March 1979 to actively seek com·ments and ideas on rulemaking for alternative sites.
Representatives from industry, State and Federal
government, public interest groups and others participated. Utilizing public comments and the results of
the workshop, the staff drafted proposed amendments
to 10 CFR Part 51 which pertain to the evaluation ofalternative sites. These amendments were submitted to

the Commission in July 1979 for their consideration.
The results of the staff deliberations on genericrulemaking and public comments received in responseto the Federal Register notification of December 14,
1978 as well as the March workshop have alreadyyielded benefits in staff review practices and the revision of environmental standard review plans to deal
more effectively with alternative siting issues.

Siting Policy Task ForceThe essential elements of nuclear power plant siting
policy are derived from the Atomic Energy Act of 1954
and are contained in 10 CFR Part 50, "Domestic
Licensing of Production and Utilization Facilities,"and in 10 CFR Part 100, "Reactor Site Criteria."
These regulations were promulgated by the Atomic
Energy Commission in 1962 and have remained essentially unchanged since that time. The authors of Part
100 recognized that experienc.e with siting nuclear
power plants was at that time limited and, in anticipa~tion of subsequent changes as experience was gained,
included in Paragraph 100.1 the statement that:"(b) Insufficient experience has been accumulated topermit the writing of detailed standards thatwould provide a quantitative correlation of all
factors significant to the question of acceptabilityof reactor sites. This part is intended as an interim guide to identify a number of factors considered by the Commission in the evaluation ofreactor sites and the general criteria used at thistime as guides in approving or disapproving proposed sites."
In the time since Part 100 was promulgated, the
NRC has issued additional siting-related 'pronouncements in the form of siting decisions on specificcases, General Design Criteria, Regulatory Guides,
Standard Review Plans, Licensing and Appeals Board
decisions, and advice from the Advisory Committee on
Reactor Safeguards (ACRS). All of these sources havecontributed to the formulation of the Commission's
current siting policy and practice. During this evolutionary period, the nuclear industry experienced arapid expansion, the use of nuclear power plants
became commonplace, and the size of such plants increased significantly. As a consequence of this expansion, some in staff practice and in the implementationof the siting regulations have evolved. In addition, theCommission's implementation of the National Environmental Policy Act of 1969 (NEPA) has added new
dimensions to siting policy.
In August 1978, the Nuclear Regulatory Commission directed the staff to develop a general policy statement on nuclear power reactor siting and a Task Force
was formed for the purpose. The Report oj the Siting
Policy Task Force (NUREG-0625) was issued in

==================================================109
August 1979. The report provides a review of current
NRC policy and practice and recommends a numberof changes to achieve the following goals:• To strengthen siting as a factor in defense-indepth by establishing requirements for site approval that are independent of plant design consideration. The present policy of permitting plantdesign features to compensate for unfavorable sitecharacteristics has resulted in improved designs
but has tended to deemphasize site isolation.
• To take into consideration in siting the risk
associated with accidents beyond the design basis(Class 9) by establishing population density anddistribution criteria. Plant design improvements
have reduced the probability and consequences ofdesign basis accidents, but there remains theresidual risk from accidents not considered in thedesign basis. Although this risk cannot be completely reduced to zero, it can be reduced by
selective siting.
• To require that sites selected will minimize therisk from energy generation. The selected sites
should be among the best available in the region
where new generating capacity is needed. Siting
requirements should be stringent enough to limit
the residual risk of reactor operation but not sostringent as to eliminate the nuclear option fromlarge regions of the country. This is because
energy generation from any source has its
associated risk, with risks from some energy
sources being greater than that of the nuclear option.
Nine changes were recommended by the Siting
Policy Task Force for consideration by the Commission (NUREG-0625, pp. 46-63):(1) Revise Part 100 to change the way protection isprovided from accidents by incorporating fixed exclusion and protective action distances andpopulation density and distribution criteria.
(i) Specify a fixed minimum exclusion
distance based on limiting the individual
risk from design basis accidents. Furthermore, the regulations should clarify therequired control by the utility over activities taking place in land and water
portions of the exclusion area.
(ii) Specify a fixed minimum emergency
planning distance of 10 miles. The
physical characteristics of the emergency
planning zone should provide reasonableassurance that evacuation of persons, including transients, would be feasible ifneeded to mitigate the consequences ofaccidents.
(iii) Incorporate specific population densityand distribution limits outside the exclu-

sion area that are dependent on the
average population of the region.
(iv) Remove the requirement to calculate
radiation doses as a means of establishingminimum exclusion distances and low
population zones.
(2) Revise Part 100 to require consideration of thepotential hazards posed by man-made activities
and natural characteristics of sites by establishingminimum standoff distances for:(i) Major or commercial airports(ii) LNG terminals(iii) Large propane pipelines(iv) Large natural gas pipelines(v) Large quantities of explosive or toxic
materials(vi) Major dams(vii) Capable faults.
(3) Revise Part 100 by requiring a reasonableassurance that interdictive measures are possible tolimit groundwater contamination resulting fromClass 9 accidents within the immediate vicinity ofthe site.
(4) Revise Appendix A to 10 CFR 100 to better reflect
the evolving technology in assessing seismic
hazards.
(5) Revise Part 100 to include consideration of postlicensing changes in offsite activities:(i) The NRC staff shall inform local
authorities (planning commission, county commissions, etc.) that control activities within the emergency planningzone (EPZ) are the basis for determining
the acceptability of a site.
(ii) The NRC staff shall notify those Federal
agencies, as in Item (i) above, that may
reasonably initiate a future Federal action that may influence the nuclear
power plant.
(iii) The NRC staff shall require applicants tomonitor and report potentially adverse
off-site developments.
(iv) If, in spite of the actions described in
Items (i) through (iii), there are off-site
developments that have the potential forsignificantly increasing the risk to the
public, the NRC staff will consider
restrictions on a case-by-case basis.

(6) Continue the current approach toward site selection from a safety viewpoint, but select sites so that
there are no unfavorable characteristics requiring
unique or unusual design to compensate for site inadequecies.
(7) Revise Part 100 to specify that site approval beestablished at the earliest decision point in the110review and to provide criteria that would have to
be satisfied for this approach to be subsequently
reopened in the licensing process.
(8) Revise Part 51 to provide that a final decision
disapproving a proposed site by a State agency
whose approval is fundamental to the project
would be a sufficient basis for NRC to terminate
review. Such termination of a review would thenbe reviewed by the Commission.
(9) Develop common bases for comparing the risks forall external events.

Early Site Reviews
Utilities are continuing to use the early site review
process adopted by the NRC in 1977 to improve reactor licensing. Two additional applications have beentendered under these procedures-the Carroll County
Station (Ill.) and the Fulton Station (Pa.). The Fulton
application is an amendment to the utility's previous
application for construction permits. In addition, thestaff has completed site environmental and safety
documents for the North Coast (Puerto Rico) application and submitted these documents to the Atomic
Safety and Licensing Board and the Advisory Committee on Reactor Safeguards. The review is being
delayed awaiting notification from the utility as towhether it wishes NRC to continue the review. Threeother applications are in various stages under the earlysite procedures-Blue Hills, Texas; Douglas Point,
Maryland; and Fort Calhoun, Nebraska.

Environmental Standard Review Plans
Environmental standard review plans (ESRPs) constitute a series of instructions developed for the NRC
staff's environmental review of applications fornuclear power plant construction permits. Their main
purpose is to improve the quality of staff reviews of environmental issues. The plans also provide guidance toapplicants regarding the information and criteria considered essential to the staff's environmental review
process. The ESRPs, 93 in number, were issued
throughout 1977 for draft review and public comment
as NUREG-0158, Parts I, II and III. In May 1979 therevised Environmental Standard Review Plans were
issued as NUREG-0555. As internal procedures andpositions or Commission policies change, the ESRPs
will be modified to keep them current with thesechanges.

Environmental Impact Statements
During 1979, the staff had undertaken to revise theformat used for the Commission's environmental impact statements for the construction and operation of
nuclear power plants. This effort was performed
within the framework of the requirement to revise theCommission's regulations covering licensing andregulatory policy and procedures. Such revision wasundertaken in compliance with new regulations
published by the Council on Environmental Quality(CEQ) in the National Enviromental Policy Act: Implementation oj Procedural Provisions (Federal
Register, Vol. 43, No. 230, November 29, 1978).

Social and Economic Issues
Forecasting Socioeconomic Impacts. Hearing issues
on s~ci~economic i~pacts have led to a heightened appreCIatIon of the Importance of empirical studies ofthese impacts at regionally and environmentally
diverse locations of nuclear power plants as an aid toimproving the analytical basis for forecasting sllch impacts in new licensing actions. The first of these
retrospective studies analyzed socioeconomic impacts
on the communities surrounding the Pilgrim I Nuclear
Station (Massachusetts) and the Millstone I and II
Nuclear Station (Connecticut). Performed for the
NRC under contract by the Oak Ridge National
Laboratory (ORNL), it was issued in September 1977
under the title, "A Post Licensing Study of Community
Effects At Two Operating Nuclear Power Plants"(ORNL/NUREG/TM-22). A continuation of this typeof post-licensing study of socioeconomic impacts has
led to two additional NRC-funded studies the first
focusing on the nuclear plant sites at Bruns~ick 1 and
2 (North Carolina) and Hatch 1 and 2 (Georgia), andthe second at the Trojan plant (Oregon):(1) "Socioeconomic Impacts of Nuclear Power Plants:
A Paired Comparison of Operating Facilities"(NUREG/CR-0916), ORNL, July 1979.
(2) "Social and Economic Impacts of the Trojan
Nuclear Power Plant: A Confirmatory Technology
Assessment" (NUREG/CR-0973), University ofWashington, Program in the Social Managementof Technology, October 1979.
A socioeconomic study of specialized scope was performed for the NRC by the Pennsylvania State University on the" Effects of Nuclear Power Plants on Community Growth and Residential Property Values"(NUREG/CR-0454). Issued in April 1979, this study
concluded that the four northeastern plants (Pilgrim
Millstone, Oyster Creek, and R.E. Ginna)demonstrated no significant influence on the price of
housing and that growth rates for the years following
plant construction were higher than the period prior toconstruction.
Contract research is continuing in the developmentof analytical tools to evaluate visual esthetic impacts ofalternative closed-cycle cooling systems and to improve the forecasting of the number of incoming con-

===================================================111
struction workers, their family characteristics and probable residential location, in order to assess the likely
degree of stress on community services and housing.
A new contractual effort was initiated with Mountain West Research, Inc., in late 1978 to study thesocial and economic consequences of siting, constructing, and operating nuclear power stations in the
United States. Fourteen stations at 13 sites wen!
selected for study: Surry 1 and 2 (Va.); Three Mile
Island 1 and 2 (Pa.); Peach Bottom 2 and 3 (Pa.); Zion
1 and 2 (Ill.); Cook 1 and 2 (Mich.); Oconee 1-:::'(S.C.); Rancho Seco (Cal.); Fitzpatrick/Nine Mile
Point (N.Y.); Calvert Cliffs 1 and 2 (Md.); Crystal
River 3 (Fla.); St. Lucie 1 and 2 (Fla.); Arkansas 1 and
2 (Ark.); and Diablo Canyon 1 and 2 (Cal.). Selection
criteria prescribed that plant sizes be in excess of 800
MWe, with an expected operating period of at least 12
months; regional diversity and an appropriate spectrum of variations in the rates of population growth in
the host county were provided for in the selection, as
were plant locations at varying distances from population centers of 50,000 or more. Specific socioeconomic
effects at the local and regional level being studied include: employment, retail sales, public services, housing, public finance (especially tax benefits), community participation and conflict, and community perception of social well-being. Work on the methodology
phase of the study was completed in June 1979, anddetailed case study work was undertaken at four sites
in July 1979, with completion of the study of all 13 sites
expected by December 1980.
As would be expected, the accident at Three Mile
Island (TMI) on March 28, 1979, substantially affected the study plan underway at that time. Not only
was TMI one of the case study sites, but there was conjecture that TMI might affect the way in which otherstations would be evaluated by local residents. The
original design had to be modified, therefore, toinclude four analytic time periods: siting, construction, operation, and the post-accident period. For
TMI there was yet a fifth period, the two-week period
following the accident, that must be studied. In order
to be able to document both the accident and the postaccident social and economic effects at TMI, it became
clear that primary data would have to be collectedfrom area residents. This data requirement led to theThree Mile Island Telephone Survey, which included
1500 households within 55 miles of the plant site. The
scope of the survey included: evacuation behavior; thedecision-making process regarding evacuation; theevaluation of the quality of information resources;
general attitudes about nuclear power and the community's economic and social outlook following the accident; the direct and indirect social and economic
costs of evacuation; and demographic descriptors. In
October 1979, a preliminary report, Three Mile Island
Telephone Survey (NUREG/CR-1093), was publishedwhich presented procedures and findings of thesurvey. The magnitude of community anxieties raisedby the TMI accident is evident in the report's estimatethat about 144,000 persons temporarily moved out ofthe zone within 15 miles of the plant site, travelling an
average distance of 100 miles to a total of 21 States,
mainly to stay with friends and relatives.
A separate contract study is in the planning stages
within NRC that would seek to determine the effect of
the TMI accident on property values as a function of
distance from the plant and of time through 1982, notonly for the TMI site but also for the four sites studied
in NUREG/CR.0454.
Independent Analysis of Need For Facility. Progresshas continued in 1979 in improving the analytical tools
for independent assessment by the staff of need-forfacility. In October 1978, the NRC published the contract study by the Oak Ridge National Laboratory(ORNL) on "Regional Econometric Model for
Forecasting Electricity Demand by Sector and by
State" (NUREG/ CR-0250). The sectors for which total
electrical energy demand are forecasted by the model
include residential, commercial, industrial, and other.
The model provides flexibility for deriving separate
forecasts for comparative purposes by making different scenario assumptions regarding such basic
causal factors as population growth, per capita income
and value added in manufacturing. Related contract
studies underway by the ORNL are expected to be
published in 1980 which will extend, update, and improve the model for staff reviews in dealing with avariety of controversial hearing issues associated with
the need-for-power issue. The titles of these studies
reflect the added dimensions of model improvement:• Comparison and Projection of Electricity Cost by
State.
• Econometric Model for the Disaggregation ofState-Level Electricity Demand Forecasts to theService Area.
• Peak Electricity Demand Predictions Using Hourly Variations.
An additional improvement not reflected in these
titles is the disaggregation of the industrial demandthrough the use of 2-digit Standard Industrial
Classification Codes (SICs).
Economic Comparison of Coal and Nuclear Energy
for Generating Electricity. Controversial hearing
issues in the licensing of nuclear power plants frequently involve the question as to whether coal or
nuclear energy is the more economical method ofgenerating electricity at a particular location. To improve the basis for the staff's independent analysis of
the comparative economic evaluations provided by theapplicant for a construction permit, the NRC issued a112

staff report in December 1978 on the subject, "Coaland Nuclear: A Comparison of the Cost of Generating
Baseload Electricity by Region" (NUREG-0480).
The study compares the economics of a 2400 MWe
nuclear and coal electric generating station in 10 different regions of the United States. Delivered coal costs
are the primary cause of regional generating costvariations; therefore, the regions were based on the
Department of Energy's (DOE) regions for delivered
coal costs. The capital cost for coal-fired generating
units includes the cost of sulfur removal. The
economics are based on a station beginning operation
about 1990 for an investor-owned utility.
The study is based on data inputs from numerous
sources, and it avoids the pitfalls of cost analyses basedon national averages by highlighting regional differences which-in addition to the transportation costsof coal affecting the delivered cost of coal to different
regions-include variations in coal characteristics,
and construction costs for labor and materials, as wellas labor productivity.
A computer program called CONCEPT, developedby the Oak Ridge National Laboratory, was used togenerate capital cost estimates by region.
The study results indicate that nuclear generating
units are either more economical or about equal tocoal generating costs for each of the 10 regions, evenusing an assumption of no recycle of plutonium or
uranium in the nuclear option. However, the study
results are regionally generic and applied to individual
licensing cases by an evaluation of the site-specific andother causal factors, if any, which would cause thecomparative cost estimates to depart significantly
from the generic assumptions and data inputs. For example, air quality standards at one site may affect thecost of coal-fired generating units differently than
nuclear; also seismic conditions at another site may affect nuclear generating costs differently than coal.
Moreover, given that forecasting is an imprecise art,
the methodology provides flexibility in facilitating thecomputation of other forecasts based on assumptionsof cost-related inputs different from those used in the
study. Among the latter are such assumptions as aplant capacity factor for both coal and nuclear plantsof 65 percent, a 30-year operating life, a 10 percent
discount rate, a 5 percent general inflation rate during
the operating period, and fixed charge rates varying
from 11 to 17 percent.

Information Base for Licensing Decisions
Demographic Data for Nuclear Sites. A draft report
was issued by the NRC in December 1977 on the subject, "Demographic Statistics Pertaining to Nuclear
Power Reactor Sites" (NUREG-0348). The staff has
prepared a revised and expanded version of this docu-

ment, increasing from 104 to 145 the number of sites
treated, and including additional tables of populationand population 'center information. This improved
data for siting policy and practice was published in
October 1979 (NUREG-0348).
Data Base for Aircraft Risk Assessment. Recent
licensing hearings have experienced an increasing
scrutiny of reliability factors, such as crash densities
used to estimate aircraft-impact likelihoods fornuclear stations which have a commercial airport
within 5 miles of the site. In response to the need forreliable historical data for aircraft accidents, the NRC
has issued a report in June 1979 on "Aircraft Impact
Risk Assessmeent Data Base for Assessment of Fixed
Wing Air Carrier Impact in the Vicinity of Airports"(NUREG-0533) .
Hudson River Fish Impact Study. In support oftestimony at EPA's hearing on Hudson River power
plant cooling systems, NRC funded the developmentof an investigation resulting in a report on "Fish Protection at Steam-Electric Power Plants: Alternative
Screening Devices," pubished by Oak Ridge National
Laboratory as Report No. ORNL/TM-6472, July 1979.
This report, discussing applications and limitations ofalternative intake designs, has general applicability topower plant siting and design where potential loss ofearly life stages of fish is of concern.
Predicting Fishery Impacts. Because of the proliferation of complex mathematical models describing
fish population dynamics, NRC has contracted with
the University of Wahington College of Fisheries toundertake a comparison of existing models to prOVidethe staff with guidance for use of models in evaluating
the impact of power plants on fisheries. The first project report, "Comparison of Simulation Models Usedin Assessing the Affects of Power-Plant-Induced Mortality on Fish Populations" (NUREG/CR-0474), waspublished in October 1978. This study concludes thatexisting models are limited in usefulness for making
quantitative predictions of population impacts.
Specific deficiencies and recommendations for future
efforts of modelers are presented. A continuation ofthis study will result in a comparison of the costeffectiveness of various modeling approaches. A
related study at the Battelle Pacific Northwest
Laboratory has resulted in the publication in March
1979 of a report, "The Application of Fisheries
Management Techniques to Assessing Impacts: Task I
Report" (NUREG/CR-0572). This study was undertaken to learn whether existing fisheries management
techniques, which usually require significantly lessdata than do mathematical models, can be used to
document the impact of a power plant on a fisheryresource.

====================================================113
Generic Study on Asbestos Fibers. Because of national concern over the potential carcinogenicity ofairborne asbestos fibers, NRC sponsored a study by the
Argonne National Laboratory to determine more
precisely the basis of concern over the use of asbestos
fill material in power plant cooling towers. The final
report, "Asbestos in Cooling-Tower Waters"(NUREG/CR-0770) , was published in March 1979.
The study concluded that the concentration of fibers
found in a number of power plant effluents would notconstitute a health hazard.
Other Information on Ecological Impacts. Other
NRC studies under way which will improve the information base for assessing ecological impacts are:
• The relationship between shipworm abundanceand distribution at Barnegat Bay in New Jersey
and changes in temperature and salinity causedby the operation of the Oyster Creek Nuclear Station.
.
• The ecological significance of fish impingementon the intake screens of the Arkansas Unit One
Nuclear Station.
• The toxicity and environmental importance ofchlorine and heavy metal discharges in the effluents of nuclear power plants, the frequency
and significance of pathogenic amoebae in cooling systems, and quantification of mortality byentrained organisms in once-through condenser
cooling systems.
• The application of aerial remote sensing techniques to routine terrestrial monitoring, and the useof reconnaissance level information for evaluating
potential impacts of alternative sites.
Meteorological Measurement and Prediction. During 1979, a survey study sponsored by the NRC wascompleted by the Brookhaven National Laboratory on
the state-of-the-art in assessing atmospheric diffusion
conditions in coastal regions. The study identified
meteorological measurement programs, test conditions, and needs for additional research to avoid
underestimating concentrations in the event of accidents at reactor sites in the coastal zone.
The staff also sponsored a state-of-the-art survey ofthe transport and diffusion of hazardous materials at
the Los Alamos National Laboratory. The purpose of
the study was to identify modeling requirements ofeither buoyant or sub-buoyant plumes resulting fromreleases, including explosions, of hazardous materials.
The summary also indicated research needs.
The staff sponsored technical assistance by the
Naval Surface Weapons Center on the assessment of
the state-of-the-art regarding the potential for missiles
to become airborne in tornadoes. The principal purpose of this study was to determine whether the typesof missiles the staff routinely postulates for purposes ofassessing reactor design are adequate. The study con-cluded that several missiles specified by the staff would
be unlikely to fly in the event of a severe tornado. As aresult of these studies, the staff is reconsidering its present criteria.
Improved Interfacing with Utilities Regarding
Meteorological Data. The staff has standardized theformat for reporting meteorological data collected atreactor sites for reactor licensing. In the past, summarized data were required for consideration in reactor licensing, but the format for such information was
not specified. Improved data acquisition recording
systems in the private sector, and the need for standardization in the NRC's consideration of meteorological data, prompted the specification of a standard
format for reporting on-site meteorological data on
magnetic tape. Subsequent to the specification of thestandard format, receipt of magnetic tape from individual reactor sites has expedited evaluations by thestaff and has reduced errors in data handling.
Standardization of Meteorological Assessments for
Accidental Releases and Routine Releases. During
1979, the staff developed and promulgated computercodes for assessing meteorological conditions following
an accident and for routine releases. The publicationof these computer codes and reference to them in NRC
standards is expected to facilitate both the industry
and staffs efforts in future licensing situations.
Improved Access to Agencies' Water Data. During
1979, the staff established and implemented directcomputer access to EPA's STORET and the USGS's
W ATSTORE computer information and retrieval
systems. Both of these systems allow rapid access tosignificant water-related data collected at many locations around the country. The access to these systems
by NRC has allowed more speedy and accurate
evaluations of both safety-related and environmental
subjects.
Installation of Computer Information Retrieval
System for Environmental Data. During the past year,
a computerized document control system (known asTERA) was installed in NRC. This system will allow
the professional and administative staff to search forand retrieve NRC documents, including environmental data from the files more efficiently than before (seeChapter 14.)

Quality Assurance
The application of disciplined engineering practices
and thorough management and programmatic controls to the design, fabrication, construction, andoperation of nuclear power plants is essential to theprotection of public health and safety and of the environment. Quality Assurance (QA) prov·ides this

114

necessary discipline and control. Through a QA program that meets NRC requirements, all organizations
performing work that is important to safety are required to conduct work in a preplanned anddocumented manner; to independently verify the adequacy of completed work; to provide records that will
confirm the acceptability of work and manufactured
items; and to assure that all individuals are properly
trained and qualified to carry out their responsibilities.
Each NRC licensee is held responsible for assuringthat its nuclear power plants are built and operated
safety and in conformance with the NRC regulations.
In addition, the NRC has several specific QA responsibilities. First, it has a responsibility for developing
the criteria and guides for judging the acceptability of
nuclear power plant QA programs. Second, it has a
responsibility for reviewing the QA programs of eachlicensee and its principal contractors to assure that sufficient management and program control exist. Finally, NRC inspects selected activities to determine that
the QA programs are being implemented effectively.
Where QA programs are found deficient, the NRC
requires appropriate upgrading. In those cases where
the QA program is not being properly implemented,
the NRC uses enforcement authority as necessary toachieve proper implementation. If a generic QA problem develops, improvements in QA programs are
made industry wide.
Through the NRC topical report program, the industry has widely adopted standardized QA programs
which can be used on new projects without a newreview. As of the end of the fiscal year, a total of 38
topical reports on quality assurance from manufacturers of nuclear steam supply systems, architectengineering firms, constructors, and utilities havebeen found acceptable by the NRC and other reports
are under review.
NRC is engaged in activities, also under the topicalreport program, that are intended to minimize or
eliminate the need for redundant audits of suppliers
without reducing the confidence that work is proceeding satisfactorily in accordance with regulations.
NRC is in the process of reviewing a topical report
describing the ASME certification and inspection program which, if found acceptable, could be endorsed asa "third party" audit program. Successful achievementof this objective should further reduce the need forpre-award audits and for yearly programmatic audits
by purchasers.
In an effort to improve QA, the acceptance criteria
contained in Section 17, "Quality Assurance," of theStandard Review Plan, NUREG-75/087, which serves
as the basis for determining the acceptance of QA programs, were updated to provide additional QA controls to give further confidence in the acceptability of
QA programs.

Since TMI and other incidents, the overall structure
for determining and acceptable QA program, including the capabilities and qualifications of individuals performing quality-affecting activities, areundergoing a review and evaluation to identify areas
where further improvements can be made.

Systematic Evaluation of Operating Reactors
The Systematic Evaluation Program (SEP) staff is
responsible for the review of 11 older licensed
operating power reactors, applying current licensing
criteria, and for documenting the results-including
the need for any necessary plant changes. The major
objectives of the SEP are set forth in the 1978 NRC Annual Report, pp. 59 and 62.
Phase I of the SEP, the development of a list of
topics to be used in performing the systematic evaluations, has been completed. As a result, a comprehensive list of topics and definitions of staff saftey objectives, together with a review procedure that considers
the effect of these topics on Design Basis Events, weredeveloped. Phase II of the SEP, the actual evaluationof the eleven older facilities, was approved by theCommission in November 1977 and is now scheduled
for completion by May 1982. The original completion
date had been January 1981. The principal reasons forthe slippage is the fact that the level of effort wasunderestimated and the other, higher priority
efforts-such as response to the TMI-2 accident andequipment qualification reviews-have diverted
significant manpower from the SEP effort. Steps havebeen taken to address these concerns by establishing an
Assistant Director for SEP and by the dedication of additional manpower to the program.
Topics not applicable to a plant design or undergeneric review have been deleted from the plant topic
lists. Of the remaining topics for each plant, morethan 50 percent are in various stages of review. Thiseffort has progressed to the point where facility Design
Basis Event (DBE) reviews, which directly constituteanother 25 percent of the topics, have been started
concurrent with the review of the remaining plantspecific topics.
The DBE reviews will become the basis for determining the capability of a plant to properly respond to
postulated accident/incident scenarios and the needfor conformance to current licensing criteria. Most
topics and all DBEs will be integrated into a final
assessment for each facility to determine the overallrequirements for facility upgrading.
One of the major topics in the SEP involves seismic
design considerations. Seismic design criteria evolved
significantly during the period 1956 to 1967, during
which the 11 SEP facilities received their Construction
Permits. Consequently, the seismic designs of theseplants vary considerably.

===================================================115
The SEP facilities follow two groups based upon the
degree to which seismic design was originally considered. The licensees of the earlier SEP facilities areembarking on seismic re-evaluation programs of theirown to supplement the existing data base which is forthe most part far less rigorously developed than would
be expected today. These programs are being
developed such that they are comprehensive enough toprovide the staff with sufficient data to enable an
overall assessment of the seismic safety of thesefacilities.
The NRC staff is currently reviewing the original
seismic design documentation of the later facilities. In
some cases, the existing information has been sup.;,plemented by NRC staff studies to verify staff
judgments. All of these plants have been visited to dateby specially staffed seismic teams to gain first handknowledge of facility geometry and to visually identify
any anomalies.
In September 1978, a team of expert seismic consultants was formed to assist the staff in reviewing theplant designs. The team had completed a review ofDresden 2 and the printed version of the evaluation
report from the team on that review was pending at
the close of the report period. No major deficiencies inthe seismic design of this facility which would affectpublic health and safety have been identified, butseveral issues have been identified which will require
more detailed studies, and possibly retrofitting, toverify the adequacy of the seismic design to meet theintent of current design crit~ria.

ANTITRUST ACTIVITIESAs required by law since December 1970, the NRC
has conducted prelicensing antitrust reviews of all applications for nuclear power plants and certain othercommercial nuclear facilities. These reviews assurethat the issuance of a particular license will neither
create nor maintain a situation inconsistent with theantitrust laws. The NRC holds a hearing whenever one
is recommended by the Attorney General and also considers whether antitrust issues raised by the NRC staff
or intervenors should be subject to a hearing.
Remedies to antitrust problems usually take the formof conditions attached to licenses. Such license conditions may result either from hearings or from nonhearing negotiated settlements.
Antitrust hearings are held separately from those on
environment, health and radiological safety matters.
So that antitrust reviews do not delay NRC liCilensing
decisions, applicants are required to submit specified
antitrust information to the NRC at least nine months,
but not earlier than 36 months, before other parts ofthe construct,ion permit applications are filed for acceptance review. Additionally, NRC performs an-titrust reviews prior to issuing operating licenses todetermine whether significant changes in applicants'
activities have occurred since the construction permit
antitrust reviews which would necessitate an antitrust
hearing.
Since the inception of NRC's antitrust program, 90
initial construction permit antitrust reviews have been
performed and one is pending. Based on these reviews,
the Department of Justice recommended 17 for hearing, 24 for "no hearing" because applicants agreed toantitrust license conditions, and 49 for "no hearing,"
without need for conditions. In addition to thesereviews, NRC has reviewed and sought advice fromthe Department of Justice in 34 cases in which additional applicants are seeking part ownership participation in nuclear plants for which the initial applications
had been reviewed previously. No hearings have been
recommended for these additional applicants.
The NRC has also sought the Attorney General's advice for two applications for operating licenses where
the Commission determined that significant changes
in the applicants' activities have occurred. The Attorney General recommended hearings in both cases.
Additionally, the NRC staff has completed operating
license reviews of twelve applications in which it
found no significant changes to have occurred and iscurrently reviewing twelve others.
In its antitrust program, NRC has reviewed over
170 private, public and cooperative utilities, which accounted for approximately 84 percent of total kilowatt
hour sales in the United States in 1977. The NRC hasreviewed 72 of the top 100 utilities, ranked by kilowatt
hour sales, in the United States.
Significant developments have occurred during
fiscal year 1979 in several antitrust proceedings. Thesedevelopments include the following:• On September 6, 1979, an Atomic Safety andLicensing Appeal Board issued its decision on the
antitrust hearing conducted for the application
by five Ohio and Pennsylvania utilities to construct and operate the Davis-Besse Nuclear Power
Station, Units 1-3, and the Perry Nuclear Power
Plant, Units 1-2. The appeal board's decisionessentially affirmed the initial decision of thelicensing board and supported the position of the
NRC staff that issuance of licenses to the applicants of these facilities would tend to maintaina situation inconsistent with the antitrust laws.
The appeal board also affirmed, with somemodifications, the license conditions believed
necessary to remedy the situations inconsistent
with antitrust laws found by the licensing board.
• In June 1978, the NRC issued a Notice of Violation to the Cleveland Electric Illuminating Company (CEI) regarding non-compliance with antitrust conditions imposed on the Davis-Besse Unit

116

1 license pertaining to transmission services for
the City of Cleveland. Since that time, the NRC
has issued an order modifying the license, theCleveland Electric Illuminating Company has requested a hearing, and the Department of Justice
has requested NRC to impose civil penalties on
CEI. These matters are currently under consideration.
• On June 28, 1978, the Commission ordered an
antitrust hearing with respect to Florida Powerand Light Company's application to constructand operate the St. Lucie, Unit 2, Nuclear Power
Plant. The Commission's decision was in response
to a late petition to intervene and request a hear~ing filed by the Florida Municipal Utilities
Association and several Florida cities. During
1979, discovery procedures were initiated but
were subsequently delayed pending possible set~tlement. Settlement negotiations are continuing.
• In 1978, the Attorney General advised the Commission that "significant changes" had occurredsince the construction permit antitrust reviews forboth the South Texas and Comanche Peakfacilities. Consequently, the Attorney Generalrecommended hearings in both cases. During
fiscal year 1979, the discovery phases of both antitrust proceedings have been consolidated
because of overlapping issues and parties. Hearings are scheduled to begin in 1980.
• Discovery has been progressing in the antitrust
proceeding for the Pacific Gas and Electric Company's application for its Stanislaus nuclear
power plant.
• The Commission has put in effect and publishedin the Federal Register two changes in its antitrust
review procedures. The first is an effective rulethat reduces or eliminates the burden of reporting
antitrust information by license applicants who
own small amounts of generating capacity. The
second is a revised procedure by which the Commission has delegated to the staff the authority todetermine during the operating license antitrust
review whether, since the construction permit
antitrust review was completed, "significant
changes" have occurred in an applicant's
activities which would raise antitrust concerns.

INDEMNITY AND
FINANCIAL PROTECTION
Increase in Levels
Of Required Financial Protection
In January 1979, the two nuclear energy liability insurance pools, American Nuclear Insurers (ANI) and
Mutual Atomic Energy Liability Underwriters(MAEL U), informed the Commission that effective

January 1, 1979, the combined maximum amount ofprimary liability insurance available from the pools
would be increased from $140 million to $160 million.
Subsection 170b. of the Atomic Energy Act of 1954,
as amended, requires licensees of commercial nuclear
power plants having a rated capacity of 100,000 electrical kilowatts or more to provide proof to the Commission that they have financial protection in an
amount equal to the maximum amount of liability insurance available at reasonable cost and on reasonable
terms for private sources. In view of the increase, theCommission amended Part 140 of its regulations to increase that amount of primary financial protection required for certain reactor licensees effective May 1,
1979, to give these licensees adequate time to purchasethis insurance. In addition, in compliance with 10
CFR Part 140, those persons licensed to possess
plutonium in the amount of five kilograms or more
and persons licensed to process plutonium in theamount of one kilogram or more for use in plutonium
processing and fuel fabrication plants were also required to provide financial protection in the amount of$160 million.
Subsection 170c. of the Act provides that (a) the aggregate indemnity for all persons indemnified in connection with each nuclear incident shall not exceed$500 million, and (b) that this amount of indemnity
shall be reduced by the amount that the financial protection required exceeds $60 million. The aggregate
liability for each nuclear incident is limited to $560
million. Presently, the secondary financial protection
layer is $335 million, (i.e., 67 licensed operating
power reactors over 100 MW(e) x $5 miflion). As aresult of the increase in the pools' underwriting capacity, Government indemnity will be $65 million forfacilities required to maintain the maximum amountof financial protection, i.e, $560 million less financial
protection of $495 million ($160 million plus $335
million). Government indemnity for large power reactors will be phased out when the sum of the first andsecond layers provides liability coverage of $560
million. Under the current level of primary financial
protection required by the Commission, this will occur
when 80 commercial reactors have been licensed.
After that point, the limit of liability for a single
nuclear incident would increase without limit in increments of $5 million for each new commercial reactor licensed.

Financial Protection for
Three Mile Island Units 1 and 2
On May 1, 1979, ANI and MAELU informed theCommission and Metropolitan Edison Company,
Jersey Central Power and Light Company and Pennsylvania Electric Company, the holders of licenses

117

authorizing operation of the Three Mile Island (TMI)
Nuclear Station, Units 1 and 2, that because of the
March 28, 1979, accident at TMI, (see Chapter 2) thepools were unwilling at that time to make $160 million
in nuclear liability insurance available for the TMI
site, despite the licensee's request for such increased
coverage. The pools' principal reason for not increasing the primary insurance available (from $140
million to $160 million) for TMI was their desire tolimit clearly to $140 million their potential liability for
claims and claims expenses arising out of the March 28
accident. The pools were opposed to increasing the
primary insurance layer to $160 million because they
could not be assured that the additional $20 million
would not be used to satisfy public liability claims
associated with the March 28 accident which arise
either prior to or subsequent to May 1, 1979.
The Commission notified the licensees for TMI that
it will be necessary for them to demonstrate that they
are in compliance with NRC regulations by providing
to the Commission evidence that $160 million inprimary financial protection for both Units 1 and 2 isin place as of May 1, 1979, i.e., is effective as of that
date. At present, the primary financial protection being provided for the Three Mile Island site is $140
million. The insurance pools have proposed an endorsement, which the Commission staff has reviewed
and finds to be acceptable, that would provide $140
million in primary insurance to both Three Mile
Island, Units 1 and 2, with an additional $20 million
for Unit 1. The licensee is presently trying to obtain
additional insurance coverage of $20 million apartfrom the present policy maintained by the licensee
with the insurance pools.
On a related matter, the indemnity agreement executed by the licensee and the Commission requires
that, in the event of payments made by the insurers
under an insurance policy used as financial protection
which reduces the aggregate limit of the policy, thelicensee must apply to its insurers for reinstatement of
the amount of these payments. The licensee has requested reinstatement of the approximately $1.3
million paid out for claims and claims expenses arisingout of the March 28 accident. Insurance pools
representatives have informed the Commission staff
that they have decided not to reinstate these funds for
Unit 2, although they will reinstate them for Unit 1
through a separate supplementary insurance policy.
The practical effect of not reinstating that funds paidout for the March 28 accident is that, if there wereanother accident at Unit 2, there would not be the fullamount of primary liability insurance to pay public
liability claims resulting from such an accident.
If damages in a new accident exceed $140 million
and the secondary financial protection layer is utilized, then other power reactor licensees will make up
both the $20 million and $1.3 million claims expenses

differences through the retrospective premium assessment, by contributing at an earlier point to their shareof the damages than would be the case if the accident
had occurred at some other site with $160 million inprimary insurance. If the damages exceed bothprimary and secondary financial protection layers,then government indemnity would make up for the increment of $20 million and would be a maximum of$85 million instead of $65 million. The limitation of
liability would remain at $560 million. Total protection for the public would be unchanged.

Indemnification of Storage of Spent Fuel
At Distant Reactor Locations
In November 1977, after public notice, the Commission issued amendments to the operating licenses ofCarolina Power and 'Light Company's (CP&L)
Brunswick Steam Electric Plants, Units 1 and 2, andH. B. Robinson Steam Electric Plant, Unit 2, toauthorize CP&L to store irradiated fuel from theRobinson reactor in either of the spent fuel storage
pools at the Brunswick facility and to have this storage
indemnified. On January 8, 1979, the Commission
published a notice in the Federal Register (44 FR 1751)
requesting public comment on specific requests by twoother utilities, Duke Power Company and Commonwealth .Edison Company, to indemnify spent fuelstored at a reactor site different from the one where it
was generated, as well as the generic issue of indemnification of spent fuel generated at one reactor but
stored at another. Sixteen comments were received by
the Commission and evaluated by the staff. The questions relating to the storage of spent fuel are presently
under consideration by the staff.

Claims Handling Procedures Following
TMI
Representatives of American Nuclear Insurers (ANI)
arrived at Harrisburg, Pa., on March 29, 1979, theday after the Three Mile Island accident and began toassess the desirability of establishing a claims office.
Following the advisory by the Governor of Pennsylvania that pregnant women and pre-school aged
children living within a five mile radius of the plantshould leave the area, ANI established a claims office
to pay claims for living expenses for these people, as
well as others who had special medical problems.
On March 31, 1979, the first day of operation at the
emergency claims center, ANI made payments ofalmost $12,000. The payments increased daily andreached a peak of $167,286 on April 9, 1979. As of theend of 1979, cumulative payments for evacuation expenses and lost wages made to approximately 12,000
individuals were $1,306,495.

118

A total of over 4,200 claims were received by ANI in
1979, including 116 economic consequence claims.
Not included in the total payments were the expenses
incurred by the insurance pools, totaling approximately $160,000 at year's end.

Determination of an
Extraordinary Nuclear Occurrence
On July 23, 1979, the Nuclear Regulatory Commission published a notice in the Federal Register (44 FR
43128) that the Commission was undertaking a determination as to whether the March 28, 1979, accident
at the Three Mile Island Unit 2 reactor (TMI-2) constituted an "extraordinary nuclear occurrence" (ENO)
as defined in the NRC regulations, 10 CFR Part 140,
subsections 140.84 and 140.85. On August 17, 1979,
the Commission directed that a panel composed ofmembers of the principal staff be formed to evaluate
public comments received in connection with our July
23 notice and other technical information assembled
by the Commission from its own and other sources. Inlate December 1979, the panel completed its investigation, evaluation, and analysis and reported to theCommission its findings and recommendation. The
panel recommended that the Commission determinethat the Three Mile Island accident did not constitutean ENO as defined in the Commission's regulations.
This recommendation is advisory only. The Commission will make the final determination as to whether
the accident constitutes an ENO. If the Commission
accepts the panel's recommendation, defendants inThree Mile Island lawsuits would not be required towaive certain traditional defenses available to themand claimants would have the same rights that theynormally have under existing negligence law.

Indemnity Operations
As of September 30, 1979, 134 indemnity agreements with NRC licensees were in effect. Indemnity
fees collected by the NRC from October 1, 1978,
through September 30, 1979, totaled $1,068,175.
Total fees collected since the inception of the program
are $20,103,254. Future collection of indemnity fees
will continue to decrease as the indemnity program isphased out for commercial reactor licensees. No
payments have been made under the NRC's indemnity
agreements with licensees during the 22 years of the
program's existence.

Insurance Premium Refund
The two private nuclear energy liability insurance
pools, American Nuclear Insurers and the Mutual
Atomic Energy Liability Underwriters paid topolicyholders the thirteenth annual refund ofpremium reserves under their Industry Credit Rating

Plan. Under the plan, a portion of the annual
premiums is set aside as a resrve for either payment oflosses or ultimate return to policyholders. The amountof the reserve available for refund is determined on thebasis of loss experience of all policy holders over thepreceding 10-year period. Refunds paid in 1979
totaled, $2,054,989, which is approximately 60 percent of all premiums paid on the nuclear liability insurance policies issued in 1969 and cover the period
1969-1978. The refunds represent 84.9 percent of thepremiums placed in reserve in 1979.

ADVISORY COMMITTEE ON
REACTOR SAFEGUARDS
The Advisory Committee on Reactor Safeguards isan independent panel of advisors statutorily established to review construction permit and operating license
applications for nuclear power reactors and othernuclear facilities and to report its findings to the
Nuclear Regulatory Commission (NRC) which are
made part of the public record. The Committee also
provides advice to the Commission on a wide varietyof safety-related issues such as the adequacy of proposed reactor safety standards, reactor safety research,specific technical issues of a topical nature, and thesafety of operating power reactors. Topical reviews
are performed by the Committee upon request by the
NRC Commissioners or upon its own initiative. Upon
request by the Department of Energy (DOE), theCommittee reviews and provides reports with regard
to the possible hazards of DOE nuclear activities andfacilities. An expansion of the Committee's statutory
responsibilities (Public Law 95-209) also requires
Committee review of the NRC's Reactor Safety
Research Program and submittal of an annual reportto the Congress regarding its adequacy.
During fiscal year 1979, the Committee reviewed
construction permit applications for two nuclear
power units and operating license applications for four
nuclear power units. The Committee also completed
its review of the request for Preliminary Design Approval for a balance-of-plant standard safety analysis(BOPSSAR) for Fluor Pioneer Services Incorporated
Balance-of-Plant design as applied to the Babcock-205
Standard Nuclear Steam Supply System (SNSSS). The
Committee took note of and took no exception to the
NRC Staff recommendations to license the Millstone
Nuclear Power Station Unit 2 to operate at an increased power level of 2700 MWt.
The Committee completed its review of proposed
operation of the DOE Fast Flux Test Facility (FFTF),
a 400 MWt sodium cooled fast reactor located at
DOE's Hanford Reservation in Benton County,
Washington. The FFTF design and its use of sodium asa coolant make this reactor considerably advanced in

=================================================119

The DOE Fast Flux Test Facility (FFTF) at Hanford,
Washington, will be an important test bed for determining thehehavior of fuels, alloys and other materials under breeder-reactor
operating conditions. Both the NRC staff and the ACRS were in-volved in safety reviews for this unique facility, under agfl!Cment
with the DOE. The FFTF was completed late in the year andexpected to go critical early in 1980.

comparison to light-water-cooled reactors (LWRs)
and, consequently, require standards and operating
limits which differ from those applied to LWRs.
At the request of the NRC, the Committee preparedand in July 1979 submitted to the Commissioners"Comments on the NRC Safety Research Program
Budget," NUREG.0603. This report included com·ments on the budget levels and program plans for the
supplemental request for fiscal year 1980 to support
research related to the accident at Three Mile Island
Nuclear Station Unit 2 (TMI-2) as well as for the fiscal
year 1981 budget. Special attention was focused on
both the short- and long-term implications of theTMI-2 accident.
As a result of the accident at Three Mile Island Unit
2 (TMI-2), the Committee met on several occasions toinquire into the event and the effects of this accident

on other reactor systems. Eleven subcommittee
meetings were held to investigate accident implications, assessments, and lessons learned, and a visit wasmade to the Babcock and Wilcox (B&W) simulator at
Lynchburg, Va., to observe simulator training forreactor operators. Several members of the Committee
visited the NRC Emergency Operations Center andTMI-2 site shortly after the accident to gain first-handinformation regarding the status of plant conditions.
As a result of this concentrated effort, the Committee
prepared and submitted to the Commission fivereports in which the Committee recommended anumber of nuclear power plant studies and changes.
The Committee also reviewed and reported to theCommission on NUREG-0578, "TMI-2 Lessons
Learned Task Force Status Report and Short-Term
Recommendations. At the request of the NRC/TMI
H

120===========================================
Special Inquiry Group, the Committee has provi.ded
information regarding significant recommendatIOns
applicable to B&W and non-B&W reactors.
During the latter part of the year, the Committee
began a discussion of the basic, underlying causes
which may have contributed to the accident at TMI-2.
This evaluation is to be continued during the forthcoming year.
In other activities, the Committee generated andprovided reports to the NRC on the following topics:• Transportation of Radioactive Materials• Combination of Dynamic Loads (Interim Report)• Status of Generic Items Relating to LWRs
• Quantitative Safety Goals• Proposed Rules on the Shipment of Spent Fuel
• Summary Comparison of Stainless Steel and Zir·
caloy Fuel Rod Cladding• Bailly Generating Station, Nuclear 1 Piling
Modifications• Pipe Cracking in Light Water Reactors• Studies to Improve Reactor Safety• Licensee Event Reports
During the fiscal year, the Committee prepared thefollowing special reports to the Congress and Congressional Oversight Committees:• The Committees Second Annual Report to theCongress, 1978 Review and Evaluation of the
Nuclear Regulatory Commission's Safety
Research Program (NUREG-0496). This reportfocused on the NRC Safety Research Programwith particular attention directed to Loss.?fCoolant Accident/Emergency Core Coohng
Systems, Fuel Behavior, Primary System Integrity Operational Safety, Advanced Reactor Safety,
E~treme External Phenomena, Radiological Effects, Waste Management, Safeguards andSecurity, Risk Assessment, and Improved Reactor
Safety.
• Report to Hon. Morris K. Udall, Chairman,
Committee on Interior and Insular Affairs, Houseof Representatives, on the use of Li~ensee Ev~nt
Reports to identify those events whICh have Implications for improved reactor safety.
Advice to the NRC was provided on 18 proposed
Regulatory Guides and Standards, which dealt with
topics such as the following:• Loose Parts Detection• Medical Evaluation of Personnel Requiring
Operator Licenses• Quality Assurance Program Requirements• Containment Isolation Provision for Fluid
Systems• Fuel Oil Systems for Standby Diesel-Generators• Design Guidance for L WR Plant Radioactive
Waste Management Systems• Design, Testing, and Maintenance Criteria for
Normal Ventilation Exhaust System Air Filtration
and Adsorption Units in LWRs
• Selection, Design, and Qualification of DieselGenerator Units Used as Standby (Onsite) Electric Power Systems
The Committee reviewed proposed amendments toAppendices to 10 CFR Part 50, Domestic Licensing of
Production and Utilization FaCilities, on thefollowing:• Fracture Toughness Requirements• Reactor Vessel Material Surveillance Program
Requirements• Modification of Emergency Core Cooling Systems
In addition, the Committee reviewed plans to implement 40 CFR 190, "Environmental Radiation Protection for Nuclear Power Operation."
Many other safety-related matters concerning
nuclear power facility operations were considered by
the Committee. Some of these are as follows:• Upgraded seismic design bases• Anticipated transients without scram (ATWS)• Seismic design for nuclear power plant piping• Use of probabilistic assessment in the licensing
process
The Committee also began activities during the
fiscal year related to the probabilistic assessment of
selected nuclear power plant incidents, the development of quantitative safety criteria for nuclear
facilities, and development of failure rate data for
nuclear plant system~ and components. These activities will continue into the forthcoming year.
In performing the reviews and preparing the reports
referenced above, the Committee met in 12 regularand one special full-session meetings. In addition, 73
subcommittee and working group meetings were held
and a total of 7 site-facility visits were made. Comments were received from members of the public withrespect to several matters evaluated by the Committee.
Members of the Committee met in Japan withrepresentatives of Japanese nuclear safety agencies todiscuss reactor safety policy and practice and thecooperative Emergency Core Cooling System program. Members of the Committee also participated in
an NRC Staff meeting with the German Reactor Safety Commission (Gesellschaft Fur Reaktorsicherheit) to
exchange experience and knowledge in the areas ofwaste management and interim spent fuel storage.

4
Materials
Regulation
Planning for the decontamination of former
materials sites included this Kerr-McGee plantin West Chicago, Ill.

The NRC's responsibilities for regulating the possession, use and disposition of nuclear materials, and thesafeguarding of nuclear materials and facilities, arecarried out by the Office of Nuclear Material Safety
and Safeguards (NMSS) under three major programs:
the fuel cycle and material safety program, discussed
below; the safeguards program, discussed in Chapter
5; and the waste management program, discussed in
Chapter 6. Qualified States in the NRC's State
Agreements Program have assumed regulatory
a1.lthority within their borders over byproduct andsource material and small quantities of special nuclear
material, and the NRC exercises oversight responsibilities concerning these programs (see Chapter 8).
The fuel cycle and material safety program includes
licensing and other regulatory activities associatedwith (1) the purification and conversion of uranium
ore concentrates (after mining and milling) touranium hexafluoride; (2) the conversion of theuranium hexafluoride (after enrichment in
Government-owned diffusion plants) to ceramicuranium dioxide pellets and their fabrication into fuelfor light water nuclear reactors; (3) storage of "spent"
reactor fuel; (4) transportation of nuclear materials;
and (5) production and use of reactor-produced
radioisotopes. Among actions in these areas during
1979, the NRC:
• Completed 14 major licensing actions concerning
uranium fuel.
• Completed actions on more than 5,100 applications for new materials licenses, license amendments and license renewals.
• Completed reviews of terminated AEC licenses
involving source and special nuclear material
operations to assist in uncovering indications ofpossible residual contamination at former licensee
sites.
• Implemented Environmental Protection Agency

regulations limiting environmental radioactivity
from reactor fuel production plants and exposure
to the public therefrom.
• Issued the final "Generic Environmental Impact
Statement on Handling and Storage of Spent
Light Water Power Reactor Fuel"(NUREG-0575) .
• Prohibited the use of seven spent fuel shipping
casks which did not meet fabrication requirements until an assessment could be made of theirsafety performance.
• Ordered medical licensees to perform tests for excessive molybdenum-99 levels in technetium-99
produced in generators at medical facilities, andto refrain from administering technetium-99 containing excessive amounts of Mo-99.
• Took a number of steps, in conjunction with the
Department of Transportation, to strengthen
regulation and inspections of shipments of lowlevel radioactive wastes to the three commercial
burial facilities in the States of South Carolina,
Nevada and Washington.

. Fuel Cycle Activities
Analyses of all light water reactor fuel production
plants have been completed in order to implement
EPA regulations (40 CFR Part 190, effective
December 1, 1979) requiring control of environmental
radioactivity so that the maximum potential exposure
af any member of the public will be limited to 25
millirems. Each license will be conditioned to limit
radioactivity in effluents to meet the requirements.
NRC has completed environmental impact appraisals for all uranium fuel cycle production plants. Suchappraisals are updated before any license is renewed,

122

URANIUM FUEL CYCLE SURVEY

OPEN PIT MINING OF URANIUM. In this view of an open pit
mine, the shovel and high-lift in the foreground remove overburden (usually crumbly sandstone).to expose the uranium ore.
The pit in the background has reached the depth where theuranium ore can be identified by scanning with a Geiger Counter.
It is then marked with small red £lags for excavation and hauling.

At this mine-mill complex, the short hauling distance from mine tomill is an added advantage. Open pit mines sometimes reach
depths of 400 feet.

and they are also prepared before issuance of majoramendments to existing licenses if any significant environmental impact can be anticipated.

In an effort to avoid repetitive analyses of the fuelcycle in each light water power reactor license proceeding, the Commission in 1974 provided a summaryof uranium fuel cycle environmental effects to be expected in support of a typical 1,000-MWe reactor.
Estimates of the environmental impact values to be
ascribed to an individual reactor were stated in theregulation 10 CFR 51.20, Table S-3. The Commission
noted that the impact values would be reexaminedfrom time to time to accommodate new technology
and information. Several proceedings, NRC actions
and litigation over the matter have taken place sincethat time. (See 1978 NRC Annual Report, pp. 76-78
for background; also see Chapter 13 of this Annual
Report under "Commission Decisions.")
In July 1979 the Commission adopted a revised
Table S-3 as a final rule to replace an interim Table
S-3 that had been promulgated in March 1977. The
final rule, which became effective on September 4,
1979, was the result of extensive rulemaking that included public hearings before a three-member board
and oral presentations to the Commission. However,consideration was limited to the environmental effectsof spent fuel reprocessing and waste management.
Radon and technetium releases are not now includedin the table, and the amount and significance of suchreleases are subject to litigation in individual reactor
licensing proceedings.
The NRC staff is studying the impact of the entire
fuel cycle in a program to revise and update fuel cycle
impact values. A narrative explanation of the Table
S-3 summary will include discussions of effects on
public health, socioeconomic conditions in fuel cycle
plant areas, and the potential risks from radioactivity
as long as it endures.
Improved Radon Estimates. NRC has initiatedseveral research projects on measurement of radonfrom uranium mines and mills. Results were publishedin 1979 by Battelle Pacific Northwest Laboratory,
Argonne National Laboratory, Oak Ridge National
Laboratory, and Ford, Bacon and Davis Utah, Inc.
The data have been used in the draft Generic Environmental Impact Statement on Uranium Milling,
and in developing a new environmental impact assessment of radon releases from uranium mining and milling. A technical report on radon impacts is being
prepared as the basis for amending the radon value inTable S-3. It will be published in 1980 in advance ofproceedings to amend the rule.
Appeal Board Hearing on Radon. Pending amendment of Table S-3, the NRC staff has presented
testimony on the new data and revised estimates forradon in 17 individual nuclear power plant licenseproceedings. The different licensing boards involved
all concluded that the environmental and humanhealth effects of radon were not Significant.

123

tinues in proposals for off-site facilities dedicated tospent fuel storage.
During fiscal year 1979, NRC approved expansionof storage capacity for seven pools at reactor sites, andapplications for expanding 12 others were pending at
year-end.

Final Environmental Statement

Uranium ore and traces of uranium in topsoils or overburdens
associated with mining emit the naturally radioactive gasradon-222. Equipment such as this continuous radon emission
monitor, designed and built as part of an NRC research programby the Argonne National Laboratory, is used to measure radon
ground flux and air concentrations-data essential to the protection of miners.

The final "Generic Environmental Impact Statement on Handling and Storage of Spent Light Water
Power Reactor Fuel" (NUREG-0575), issued in August
1979, confirmed findings of the March 1978 draft
statement that spent reactor fuel generated throughthe year 2000 can be stored in a safe and environmentally sound manner by modifying pools at reactor sites
or by providing independent storage facilities either at
the sites or away from them. The statement discussed
and took into account extensive public comments
received in response to the draft document, and supported issuance of a rule to specifically cover storage at
independent spent fuel installations.

Movements Between Reactors
There were 17 nuclear power plant proceedings
before appeal boards in 1978 when the Commission
deleted the radon-222 value from Table 3 and opened
this issue for litigation. Intervenors who had been participating requested reopening of the radon issue infour cases. By mutual agreement among the parties, asingle appeal board hearing on the radon issue isscheduled to be conducted in February 1980.
Updating Fuel Cycle Rule. In addition to amending
specific portions of the fuel cycle rule, the NRC isworking toward updating of the entire environmental
survey. A draft report of the updated "Environmental
Survey of the Uranium Fuel Cycle" was nearing completion at the end of 1979. It will examine in detail theenvironmental impacts of the fuel cycle from uranium
mining through waste disposal and the decontamination and decommissioning of nuclear facilities. The
study will concentrate on the "once-through" fuel cycle with no reprocessing of spent fuel and no recycle ofplutonium. Alternatives will be considered, however,so that if reprocessing should be initiated, with attendant recycle of uranium and solidification anddisposal of high-level reprocessing wastes, the environmental effects can be taken into account if theyoccur during the lifetime of reactors under licensing
consideration.

An application by Duke Power Company for thetransfer of spent fuel from its operating Oconee
Nuclear Station, Seneca, S. C., to the utility's
McGuire Nuclear Station in North Carolina, which
has not yet received an operating license, is an example
of methods sought to deal with the shortfall in storage
space.
The NRC staff completed safety and environmental
reviews of Duke's application during 1979. Approval
would require amendment of the McGuire plant's
special nuclear material license and authorize receipt
and storage of spent fuel generated at the Oconee station. It would be the second licensing action of its
kind, approval having been given in 1978 to Carolina
Power & Light Company for the intersite transfer ofspent fuel from its H. B. Robinson Plant Unit 2 to its
Brunswick Station. (See 1978 NRC Annual Report, p.
75.)
The staffs Environmental Impact Appraisal and
Safety Evaluation Report were issued in December
1978 and January 1979, respectively. The action has
been contested by two intervenors~Carolina Environmental Study Group and Natural Resources
Defense Council-and an evidentiary hearing was inprogress before an Atomic Safety and Licensing Boardat year-end.

SPENT FUEL STORAGE ACTIONS

Away-from-reactor Storage

The need for storage of spent nuclear fuel continues
to stimulate actions by nuclear power plant licensees
to increase capacities of storage pools at reactor sites
and to ship irradiated fuel from sites with filled pools
to others where room is available. Interest also con-

An Atomic Safety and Licensing Board has be~n appointed to conduct a hearing on the General ElectricCompany's application for renewal of its license forthe receipt and storage of spent fuel at its Midwest
Fuel Recovery Plant at Morris, Illinois. Petitions for124leave to intervene were filed by the Illinois attorney
general and a group of individuals.
The staff extended its review of a topical report by
Stone & Webster Engineering Corporation, "Independent Spent Fuel Storage Facility," containing a design
for a standard facility to be located on the site of aparent facility such as a nuclear power station. A letterof approval for the conceptual design had been issuedin July 1978; a second letter of approval was issued in
January 1979 permitting specific sections of the topicalreport to be referenced in any future site-specific application. The standard installation visualized could
store up to 1,300 metric tons of uranium oxide as spent
fuel, an amount equivalent to the volume of spent fuelthat would be discharged during about 35 years ofoperation of a 1,000-MWe nuclear power station.

COVER

SAFETYCURTAIN~~J1'------ NEW FUE L

STORAGERACKSPENT FUELSTORAGE RACKDRYSTORAGEVAULT

This is typical s~nt fuel storage pool for a light water reactor
power plant. Pools are designed to particular plant configurations,but generally are box-like colltainments about 100' long by 50'

wide by 40' deep. They are constructed of reinforced concrete,
lined with stainless steel. Diagram shown depicts a fuel storage
pool at the Browns Ferry Nuclear Plant in Alabama.

ADV ANCED FUEL ACTIVITIESReview of Plutonium Plantsof adverse natural phenomena and to protect thehealth and safety of the public (see 1978 NRC Annual
Report, p. 75).

The NRC staff made substantial progress during
1979 in evaluating the integrity and safety of six
plutonium processing and fuel fabrication plants
which are licensed to possess and process fivekilograms or more of unencapsulated plutonium. The
objective is to improve, to the extent practicable, thecapabilities of these facilities to withstand the effects

Two of the six plants have been completely analyzed
and summary documents issued which describe the effects of damage to the facilities from natural
phenomena. (NUREG-0547, regarding the Babcock &Wilcox facility at Parks Township, Pa. , and
NUREG-0621, concerning Westinghouse's facility at
Cheswick, Pa.) The analyses of plant capability in-

=================================================125
elude site characterization with regard toseismology/ geology, surface hydrology, normal andsevere meteorology, and the structural capacity towithstand severe seismic and meteorologic events.
Analysis of risk to the public involves source term
estimation, meteorological dispersion, demography,
ecology, and radiological impact. Analyses of the remaining four plants were in varying stages of completion at year-end.
Babcock & Wilcox License Renewal. During theyear, NRC renewed for a five-year period the license
for continued operation by Babcock & Wilcox Company at its Parks Township, Pa., facility, where mixed
oxide fuel pins are being manufactured for use in theFast Flux Test Facility at Richland, Wash. The staff
determined that adequate protection is provided for
employees and the public after evaluation of administrative procedures, management commitments
and engineered safety features, potential impact on
the environment, and capability of the plant to withstand severe natural phenomena. As a result of thereview, B&W improved certain ventilation and fire
safety features. License conditions require decontamination of the facility at the end of its useful life,
and the inclusion of a B&W financial commitment tocarry that out.
Decommissioning Activities. During 1979, the Babcock & Wilcox Company removed all contaminatedequipment from its high.enriched uranium fuelfabrication plant at Parks Township, Pa., and completed other decontamination requirements. Finaldecommissioning and release of the site forunrestricted use will be completed when B&W determines the future use of the facility.
The Kerr-McGee Nuclear Corporation has begun todecommission its plutonium fuel fabrication plant at
Cimarron, Okla., which is in a shutdown, standby
condition. Current work involves dismantlement of
the solvent extraction equipment used in liqUid scrap
recovery; license amendments will be needed to permit further decommissioning efforts.
At year's end, Westinghouse announced plans todecontaminate and decommission its plutonium fuelfabrication facilities, and other firms were considering
similar actions.

OTHER FUEL CYCLE ACTIVITIES
Evaluating Sites for Radioactivity
The NRC continues to be active in evaluating sitesof former radioactive material operations in order that
corrective action can be taken wherever required toprotect the public.
Formerly Licensed Sites. In response to an earlier
General Accounting Office inquiry concerning potential radiation safety problems at sites previously
operated under AEC licenses, the NRC has been ex-

An NRC Region III radiation specialist perfonns a radiation survey
in connection with the decontamination and dismantling of the

Kerr-McGee thorium plant at West Chicago, Illinois. Such
radiation surveys are used in detennining alternative methods ofdecommissioning.

amining the files of licenses terminated before 1965 toascertain that proper decontamination has been carried out.
The Oak Ridge National Laboratory has completed
for NRC the evaluation of docket files for old sourceand special nuclear material licenses, and identified
approximately 225 sites which require further evaluation. Further investigations of these sites will be carried out by either NRC or the Agreement States where
involved. The next step in this program will be theevaluation of docket files for byproduct material
licenses issued before 1965.
Surveys of Burial Sites. NRC is planning forradiological surveys at the Westlake landfill, St. LouisCounty, Missouri, and the Reed-Keppler Park, West
Chicago, Ill., where radioactive materials were buriedin the past. The surveys will define the location andquantities of materials present in order that corrective
actions can be taken.
Kerr-McGee Site, West Chicago. The Kerr-McGee
Corporation has submitted a plan for removal of factory buildings, decontamination of the· site, andstabilization of the ore residue area at a thorium andrare-earth compound production plant formerly
operated by Kerr-McGee and predecessor companies
at West Chicago, 111. Production operations ceased in
1973, .and the facility is retained under a possessiononly license. The licensee's plan is being reviewed byFederal, State and local authorities, and the NRC is
preparing an environmental impact statement.

West Valley, N. Y., Facility
The future of the West Valley, N. Y., site of the
shutdown spent fuel reprocessing plant formerly
operated by Nuclear Fuel Services, Inc., remained to
be settled at the end of fiscal year 1979. A Congres-126

sionally mandated study by the Department of Energyof options for the future of the site and/or allocation of
responsibility among the Federal and State governments and NFS has been issued to the public for comment and submitted to the Congress for consideration.
Formal agreements and decisions on the disposition of
the site have not been made, although somepreliminary discussions between Federal and State officials have begun.
One of the key aspects of deliberations on whatshould be done involves the disposition of the highlevel liquid wastes presently stored in undergroundtanks. The NRC staff has been assessing the safety ofcontinued storage of these liquid wastes. As part of thisassessment, a test of a level detection instrument showed that the pan under the tank containing the high~level, neutralized waste is defective. The same test
confirmed that the tank itself is not leaking. The loss ofthe pan as a secondary collection point for any tankleakage that might occur is undesirable, andunderscores the need for commencement of dedicatedwork to remove, solidify, and dispose of the high-levelliquid wastes at West Valley. The staff, through its
contractors, is continuing its program to assess thesafety conditions associated with this storage.
The NRC staff has continued its assessment of the effects of severe natural phenomena on the dormant
reprocessing facility. Analysis of the effects of a severe
earthquake on the fuel receipt and storage pool andthe carbon steel high-level liquid waste tanks has
demonstrated that there is no undue risk to the healthand safety of the public from the effects of an earthquake. Analysis of the effects of tornadoes on theseparations plant is in progress.

Transportation of Radioactive
Materials
Transportation of radioactive materials is regulated
at the Federal level mainly by the NRC and theDepartment of Transportation (DOT). NRC sets thestandards for "Type B" packages (those whose content
of radioactive materials requires that they be safely retained in their containers under both normal and accident conditions) and for packages containing fissile
material. NRC also makes independent evaluations of
package designs submitted by applicants and serves asa technical advisor to DOT. The roles of the DOT and
the NRC were redefined in a memorandum of
understanding signed by the two agencies in June
1979, and published in the Federal Register on July 2
1979.
'
Package designs used by contractors for the Department of Energy (DOE) are reviewed and approved bythat agency; however, the NRC has been reviewing
such package designs on a continuing basis. These
NRC reviews are not binding on the DOE, but an, NRC approval permits commercial licensees to use
these packages.

Low-Level Waste Shipments
In the summer of 1979, attention was called to anumber of instances in which packages of low-level
waste were not in compliance with Federal requirements on arrival at one or more of the three commercial burial facilities in the country. Such items of noncompliance included a fire on a truck carrying combustible waste, leaking packages, truck contaminationfrom improperly closed packages, free liquid in
packages of supposedly dry solid material, inadequately labeled packages, and improperly documented
shipments. While none of these items of noncompliance by itself represented a significant healthconcern, collectively they showed a lack of proper attention to Federal requirements for packaging andshipping of radioactive waste materials.
The Governors of the Agreement States of South
Carolina, Washington, and Nevada, in which thecommercial burial facilities are located, notified the
NRC of the repeated disregard for the Federaltransportation rules and at various times closed orlimited these facilities to certain shippers. The NRC,
in conjunction with the DOT, determined that theFederal Government should improve its assurance that
Federal regulations governing such shipments are met
and took several steps:• The NRC changed its regulations to specificallysubject its licensees to DOT requirements, andthus effectively increase the Federal inspection
capability.
• The NRC issued bulletins to all licensees to (1) inform them of the transportation incidents that occurred, of requirements for transportation of lowlevel radioactive waste materials, and of burialsite requirements, and (2) direct licensees to submit written management-approved proceduresfor the safe transfer, packaging, and transportation of these materials.
• The NRC increased its inspections at shipper andreceiver sites.
• The NRC modified its enforcement criteria to increase penalties.
• The NRC and DOT are jointly investigating ways
to improve the safety of low specific activity
material packages.
• The NRC is acquiring support from the Society of
Nuclear Medicine to improve medical waste
packages and from the Atomic Industrial Forum
to improve industrial waste packages.
• The NRC and DOT are making an effort to better
inform shippers of requirements.
• The NRC is developing a primary draft regulation for burial of low-level wastes.

==================================================127

Cb)

(e)

Low Level Walle Burial(a) Low level waste containers are dumped into burial trench andquickly covered as portions of the trench are filled. Instrumented
pipes are emplaced along deep edge of trench to measure theradioactivity of any underground seepage which might occur.
Trench bottom is sloped toward instrument side. (b) Instrument
pipe protrudes above covered trench as burial nears completion.

The American National Standards Institute Subcommittee on Transportation of Radioactive Waste,with NRC participation, is preparing a standard forthe packaging for transportation of liquid aqueousradioactive wastes from nuclear power plants. It will
require that liquid wastes be solidified prior to shipment and that a high-integrity container be used.
In March 1979, the NRC published a draft report,
NUREG-0535, "Review and Assessment of Package
Requirements (Yellowcake) and Emergency Response
to Transportation Accidents," The report was
prepared by an NRC/DOT task force following a truck
accident in September 1977 in which a shipment of
uranium concentrate (yellowcake) was spilled onto ahighway near Springfield, Colo. The draft report has
been issued for public comment, which will be considered in preparing the final report. (See 1978 NRC
Annual Report, p. 81.)(c) Temporary markers are used following burial and grassing of
covered area (instrument pipes again are visible at right). (d) A
permanent marker specifying dimensions of trench and its contents
is emplaced when earth cover is settled and grassed over. This
marker at a site in South Carolina describes a pit of 493 feet by 47
feel. Depth of such trenches depends on the medium of burial,
depth of water table, etc., but on average, are between 20 and 30
feet deep.

Irradiated Fuel Packaging
Spent (irradiated) nuclear fuel is transported off-sitein specially designed shipping casks that are capable ofcontaining the radioactive fuel assembly materials
during normal and postulated design accident
transportation conditions.
On April 6, 1979, NRC issued an Order to Show
Cause (immediately effective) prohibiting the use of
the Model No. NFS-4 packaging until a determination
is made that it meets specified requirements. During ameeting on March 29, 1979, and later by letter dated
April 2, 1979, Nuclear Assurance Corporation informed the NRC staff that a cask had not been fabricated inaccordance with its certificate of compliance. In viewof the unknown safety implications for this and otherpackages fabricated to the same design, and in the interest of public health and safety, all seven packages

128

were removed from use until an assessment could bemade of their safety as fabricated. On Dec. 12, NRC
permitted three casks to be returned to service. Eigh~teen owner/users were affected by the orders.
The staff is reviewing two applications for spent fuelcasks designed for shipment by rail that are significantly different from design concepts presently used.
Each of these designs uses a thick, solid, cylindricalcarbon steel containment vessel wall as a gamma
shield. Existing designs use lead or uranium in the
gamma shield. One factor being considered in thereview is fracture toughness of thick steel forgings. The
two spent fuel casts are the Transnuclear Inc. Model
No. TN-12 and the Nuclear Assurance Corporation
Model No. NAC-3K.

Safety of Transportation Workers(3) Observe procedures in use at carrier's facilities
and prepare suggestions for techniques to.
reduce exposure.
(4) Identify relationship between quantity ofradioactive material handled and exposure.
The study began in July 1979 and should be completed by October 1980. The information will be used
to prepare a recommendation to DOT on what further
measures may be necessary to control radiation exposures in selected portions of the transportation industry.
During fiscal year 1979, the NRC-sponsored State
Surveillance Program on Transportation of Radioactive Materials was continued. (See Chapter 8, "State
Programs," for details of the program's activities during 1979.)

In June 1979, the NRC initiated a study entitled"Radiation Exposure of Transportation Workers
Handling Large Numbers of Radioactive Material
Packages." The objectives of the study are to:(1) Identify carriers where employees may receive
exposures exceeding regulatory limits.
(2) Determine actual exposures received either bydirect measurement or by studying the carrier's
records.

Transportation in Urban Areas

Canisters such as these (shown at left) are used in "cleaning"
contaminated water at nuclear plants. The steel containers holdresins that create an ion-exchange process as radioactive liquids areflushed through. When the resins are no longer clean, or when aprescribed level of radioactivity is reached, the canisters are placed
in sealed cylinders (on flatbed truck/trailer at right) for transport
to waste burial sites.

During 1979, Sandia Laboratories, under contract
to the NRC, continued its work to assess the environmental impacts resulting from the transportationof radioactive materials through urban areas. The
study has been examining the impacts resulting fromincident-free transport, vehicular accidents duringtransport, and from other abnormal situations. In performing this study, Sandia has developed computer

==================================================129
models to account for the special features of the urban
environment. The study will form the basis for ageneric environmental impact statement, to be
published by the NRC, on the transportation ofradioactive material in urban areas.
Information resulting from the study suggests thatthe sabotage of spent fuel shipments has the potential
for producing serious radiological consequences inareas of high population density. This information, inpart, led the NRC in June 1979 to establish interim requirements for the protection of spent fuel in transit.
(See Chapter 5, "Domestic Safeguards.")

Emergency Response Planning
In October 1978, Indiana University, under contract to the NRC, began work on a study entitled"Survey of Current State Radiological Emergency
Response Capabilities for Transportation Related Incidents." The objective is to assemble and condense
available information on current state emergency
response capabilities for transportation-related
radiological incidents in order to assist NRC in its role
regarding radiological incident planning, emergency
response training, and other assistance activities
within State and local governments.
The NRC will use the information obtained fromthis study to (1) identify response requirements forprotection of the health and safety of the public withregard to transportation-related radiological incidents, (2) develop and promulgate guidance to Stateand local governments in coordination with other
Federal agencies for the preparation of emergency
response plans, and (3) determine whether additional
Federal participation is required to ensure adequate
protection of the health and safety of the public withregard to transportation~related radiological incidents. The study will be completed in fiscal year
1980.

Packaging Standards
In March 1979, the NRC issued for comment
Regulatory Guide 7.9, which identifies the information to be provided in an application for the approvalof packaging for shipping Type B, large quantity, andfissile radioactive material and presents a uniform format for presenting the information. The guide assists
the applicant in preparing an application and ensuresthe completeness of the information provided. Use of a
uniform format assists the NRC staff and others inlocating the information and aids in shortening thetime needed for the review process.
In August 1979, the NRC issued for public commenta revision to 10 CFR Part 71, "Packaging of Radioactive Material for Transportation and Transportationof. Radioactive Material Under Certain Conditions."
The NRC is considering revising its regulations for thetransportation of radioactive material to make themcompatible with those of the International Atomic
Energy Agency (IAEA) and thus with those of mostmajor nuclear nations of the world. Although several
substantive changes are proposed in order to provide amore uniform degree of safety for various types ofshipments, the Commission's basic standards forradioactive material packaging would remain unchanged. The DOT is also proposing a corresponding
rule change to its hazardous materials transport
regulations.
The major changes to NRC's regulations that are being proposed are:
(1) Elimination of the system currently used to
specify the quantities of radioactive materials
permitted to be shipped in certain types of
packages. Under the present system, all radioactive materials are divided into seven transportgroups that are used as the basis for determining
the amount of those materials that can be shipped in Type A packages and the amounts thatmust be shipped in the more stringently designed, accident-resistant Type B packages. Thissystem has proved to be unduly restrictive
because less hazardous radioactive materials included in one transport group are required to bepacked in the same manner as other, morehazardous radioactive materials belonging to
the same transport group. Under the proposed
rules, the use of a Type A or Type B package
would depend on the degree of radioactivity foreach material being shipped.
(2) Establishment of two classifications of Type B
packages. This change would facilitate foreign
acceptance of U. S. export shipments by conforming package types to international standards.

International Standards
In 1983, the IAEA will issue a revision to its Safety
Series No.6, "Regulations for the Safe Transport ofRadioactive Materials." In preparation for this revision, the IAEA has asked member countries to submit
proposed changes to the regulations and to identify
areas where revision should be considered. The NRC
prepared recommended changes to the regulations
which were submitted to the IAEA by the DOT,
which serves as the U.S. competent authority on matters involving international shipments of radioactive
material. In addition, the NRC assisted the DOT in its
review of comments submitted by various privateorganizations and other government agencies.
The NRG also participated as an observer in a
meeting of the IAEA Advisory Group on Radiation
Protection and Safety Principles for Safe Transport ofRadioactive Materials. The meeting was held in July
1979 at the IAEA Headquarters in Vienna, Austria.
The purpose of the meeting was to review the principles upon which the IAEA's transport regulations are130===================================================

tn·
This radiograph of the Liberty Bell is reduced from its original sizeof seven feet by 12 Y3 feet, reportedly the largest single sheet of
X-ray film ever exposed. The exposure to a 670-curle cobalt-60
source lasted 7 1/1 hours. The famous crack in the bell is the darkirregular line at left center.

based and to provide recommendations to a revisionpanel that will meet in 1980 to begin the comprehensive review of the transport regulations.

Radioactive Material
LicensingRadioactive materials have wide use in industrial
applications, medical diagnosis and treatment, applied research and development, in the academic
fields, and in products distributed to the public. Some
8,500 materials licenses are administered by the NRC
which require the annual processing of 5,000 to 6,000
applications for new licenses, license amendments andlicense renewals. An additional 12,000 licenses are administered by 26 states which have assumed authority
over certain materials under regulatory agreementswith the NRC, as part of the Agreement States program, (see Chapter 8). The NRC licensing program isdesigned to provide reasonable assurance that the

public health and safety is adequately protected andthat applications for licenses are processed in an efficient and timely manner.
In 1979, the materials licensing function was
reorganized. The former Radioisotopes Licensing
Branch was dissolved and two new branches werecreated: the Material Licensing Branch and theMaterial Certification and Procedures Branch. The
purpose of the reorganization was twofold. It permits
the Material Licensing Branch to focus its efforts on
the review of license applications and the Material
Certification and Procedures Branch to devote its efforts to functions related to materials licensing, such as
sealed source and device evaluations, preparation ofguides for licensees and applicants, the review ofregulatory practices, and other non-casework needs.
In March 1978, the NRC initiated a pilot
regionalization licensing program to determine thefeasibility of conducting licensing activities from NRC
regional offices. The initial effort involved six of theeight States in NRC's Region III and reviews were conducted for medical institutions and industrIal firms using gauging devices. As the program progressed, alleight of the States were included in the program, andacademic institutions and industrial research anddevelopment licensees were included in those involvedin the regional licensing program. The program isscheduled for completion in early 1980 and thefeasibility of continuing or expanding the regional
licensing program will be determined.
Consumer Products. Numerous products containing
small amounts of radioactive materials are in dailyuse. These products are authorized for distribution only after careful evaluation by the NRC indicates that
there is minimal risk to the general public. Amongthese products are smoke detector devices containing
americium-241 and liquid crystal display timepieces
containing tritium. A two-year study to determine theenvironmental impact of the use of radioactive
materials in consumer products is expected to be completed in October 1980.
Gauging Devices. These devices have wide use forcontrolling density, levels, thickness, and weight ofmaterials. NRC approves their use only after evaluation of the sealed sources and the devices to determinethat the gauging devices may be used safely by individuals who have miniIJlal training and experiencein radiation safety. Due to the relatively low radiationlevels, normal use of these devices presents minimalhazard to workers and the general public.
Gas Chromatographs. These devices typically contain radioactive materials in the form of foil or plated
sources containing nickel-63 or tritium. Due to increased concern for the environment and theusefulness in measuring small amounts of materials,
gas chromatograph usage has increased dramatically===================================================131
volume determinations and kidney function
studies are examples of these) .
• Visualizing the distribution of a radioactive drug
within an organ in order to locate tumors, bloodclots, etc.
Therapeutic radiation treatment continues to be an
important element licensed by the NRC. These procedures include the use of liquid radioactive drugs totreat certain medical conditions such as the use ofiodine-I31 for treatment of hyperthyroidism. In theradiation therapy mode called brachytherapy, encapsulated or sealed radiation sources are placed directlyon or in the patient's body to treat cancer. Naturally
occurring radium sources with their many problems
are rapidly being replaced by NRC-licensed materials
such as cesium-I37 and iridium-192. Teletherapy is
another area of radiation therapy licensed by NRC. In
this method the patient is treated at a distance withradiation from a sealed radioactive source, ~suallycobalt-BO.
Currently, the most rapidly growing area in nuclear
medicine is nuclear cardiology. Using radioactive
materials, physicians are able to identify specific areasin the heart that are not receiving an adequate blood
supply, thus predicting potential heart attack victims.
Nuclear cardiology studies also enable the physicians
to locate infarcted areas and monitor healing andrecovery processes after a heart attack occurs.
This well-logging tool uses a radioactive source in taking
measurements in boreholes. The neutron source, shown in theupper part of the logging tool, activates the different earth strata
as it passes down the hole permitting identification of thecomposition of the strata.

for analyses of substances which could contain environmentally undesirable constituents.
Well Logging. The use of radioactive materials inwell logging-in search of new energy sources and or
utilization of gas and oil fields formerly thought
underproductive-has increased markedly. Most ofthese activities are being performed by large service
companies, although many small companies are also
active in the use of well logging techniques.

Nuclear Medicine
The NRC issues licenses to hospitals and physicians
for the use of radioactive materials in the diagnosis
and treatment of patients. These diagnostic procedures include both in vitro tests involving the addition of radioactive materials to laboratory samples
taken from patients and in vivo tests, the direct administration of radioactive drugs to patients. In vivo
tests are for the purpose of:• Measuring the uptake, dilution or excretion of aradioactive drug within an organ system (blood

Advisory Committee on Medical Uses of Isotopes.
NRC's Advisory Committee on Medical Uses of
Isotopes (ACMUI) consists of physicians and medical
physics specialists from the public sector and it provides NRC with advice on many medical questions.
During a December 1978 meeting, the NRC staff
recommended minimum criteria for training and experience of physicians to be licensed to perform
nuclear cardiology studies. The committee also
discussed NRC's cooperative efforts with the Society of
Nuclear Medicine to develop a medical licensee model
program designed to minimize radiation exposure topersonnel and minimize releases of radioactive
material to the environment. During the year, in accordance with NRC's policy of rotating membershipon the ACMUI, NRC formally requested nominations
from the public to replace several members. Fourwell-qualified physicians were selected from 47 persons nominated by professional societies and membersof the public.
Order on Use of Technetium-99m. Many hospitals
obtain technetium-99 (Tc-99m), a radioactive isotope
widely used in nuclear medicine, from amolybdenum-99 (Mo-99) generator. The generator is ashielded device that contains Mo-99 absorbed onto analumina column. When a sterile saline solutioit is fed
through the column, Tc-99m, a daughter product, isremoved and Mo-99 remains on the column. Very little, if any of the Mo-99 is normally removed with the132labeled phosphate compound, localizes in skeletal areas and allows
physicians to visualize pathologic bone process.

A nuclear medicine physician examines a technetium (Tc)~99m
whole body bone image. This radiophannaceutical, a Tc-99m-

Tc-99m. Mo-99 in Tc~99m serves no diagnostic purpose, and excessive levels of this isotope in the Tc-99m
results in unnecessary radiation exposure to patients.
Although it is good practice to check for the presenceof Mo-99 as a contaminant, an NRC investigation
revealed that these tests were not being routinely performed. Since several thousand generators are sold to
medical facilities each week, and one generator can
provide Tc-99m for up to 50 patients per day, there isa potential for exposure of large numbers of persons if
excessive levels of Mo~99 occur.
On March 12, 1979, NRC issued an order requiring
medical licensees to perform a test for Mo~99 and torefrain from administering Tc-99m contaminatedwith excessive levels of Mo~99. On June 6, 1979, NRC
published a proposed rule change that contains theessentials of the order and requested comments fromthe public.
Other Significant Actions. In other 1979 actionsw

having a significant impact on medical licensees, the
NRC:• Published for comment a revised medical licensing guide. Use of the revised application form
with the revised guide is making it easier to apply
for and obtain a license for the medical use ofradioisotopes.
• Issued a final rule requiring teletherapy licensees
to perform periodic full calibration and spotcheck measurements to ensure accurate determination of the amount of radiation administeredto patients.
• Issued a new rule allowing physicians greater
latitude in using approved radioactive materials
for diagnostic procedures.
• Published a medical policy statement setting forth
NRC's future role in regulating the medical usesof radioactive materials.

5
Domestic
Safeguards

Fixed site physical security at nuclear facilities
was upgraded in new NRC regulations published in November 1979.

Section 209 of the Energy Reorganization Act of 1974, as
amended, requires the Nuclear Regulatory Commission to
include in each Annual Report to Congress a chapter
describing the status of NRC's domestic safeguards program
for the protection of certain nuclear materials and facilities.
This chapter discusses safeguards provided for licensed
facilities and activities during fiscal year 1979, covering thegeneral areas of (1) scope of NRC safeguards efforts; (2)adequacy and effectiveness of safeguards; (3) safeguards
policy issues, regulatory actions, and research and technicalassistance; and (4) NRC safeguards organization andmanagement.

Scope of NRC Programs
The Atomic Energy Act of 1954 and the Energy
Reorganization Act of 1974 direct the NRC to regulate thesafeguards provided by its licensees for certain nuclear
facilities and activities. With the objective of assuring protection of the public health and safety and the national
defense and security, the NRC designs and enforces measures to deter, prevent, and respond to (1) unauthorized
possession, theft, diversion, or use of special nuclear
material; and (2) sabotage of nuclear facilities.
Safeguards for fuel cycle facilities emphasize protection
against theft or diversion of "formula quantities" of strategic
special nuclear material (SSNM), while power reactor
safeguards concentrate on protection against industrial
sabotage.
NRC safeguards regulation during 1979 covered thefollowing:• Nineteen fuel cycle facilities.
• Selected transportation activities.
• Seventy power reactors licensed for commercial operation.
• Seventy-one non-power reactors (for research, testing,training or the production of radioisotopes).
The 19 fuel cycle facilities are authorized to possess form-

ula quantities of SSNM, which includes uranium-235 (contained in uranium enriched to 20 percent or more in the
U-235 isotope), uranium-233, or plutonium. A "formulaquantity" is 5,000 grams or more of SSNM as computed by
the formula: grams(grams U-235) + 2.5 (grams U-233+ grams plutonium). Four additional fuel cycle facilities
are authorized to possess more than one "effective kilogram"of low-enriched uranium and are, therefore, subject to
NRC safeguards material control and accounting requirements. For uranium with an enrichment in the isotope
U-235 of one percent and above, "effective kilograms" iscomputed as its weight in kilograms multiplied by thesquare of its enrichment expressed as a decimal weight fraction.
The selected transportation activities mentioned above
involve shipments of spent fuel or formula quantities oflicensed SSNM, currently amounting to about 20 per
month.

Assessment of Safeguards Adequacy Policy
The fiscal year 1978 report to Congress on domestic
safeguards (NUREG-0524), which was forwarded as aseparate document, noted that the several NRC offices withsafeguards responsibilities had varying approaches tosafeguards regulation. These differences were manifested indefinitions of adequacy and also in methods of deciding
whether safeguards were adequate for the regulated fuelcycle facilities, transportation activities, power reactors andnon-power reactors.
In January 1979, the NRC Executive Director for Operations established an internal "Task Force on Safeguards
Policy" to resolve these differences and develop an integrated approach to safeguards regulation. In July 1979,
the Commission directed the staff to revise its rules in 10
CFR Part 73 to incorporate the task force's recommendation that safeguards regulatory requirements be generally
specified in terms of malevolent acts to be thwarted (Le.,theft and sabotage), rather than by types of facilities to be134

protected. To help a licensee design a safeguards system,
such m3IeVolent acts are cliaracterized as shown in Chart 1.
The rule revisions also specified that the objective ofphysical protection systems for reactors and fuel cycle
facilities is to provide high assurance protection against
radiological sabotage, theft or diversion.
In August 1979, as a further step toward eliminating differences and conflicts within NRC safeguards regulatory
programs, the Commission decided to consolidate allsafeguards review responsibilities within the Office of
Nuclear Material Safety and Safeguards. This consolidation
became effective on October 1, 1979.

The teams then judged whether the particularfacility provided high, good, fair, or poor assurance of 'protecting against a hypothesized threat. Licensee
safeguards judged to provide high assurance of protection against the hypothesized threat are considered topossess the desired level of safeguards capability.
Licensee safeguards judged to provide good assuranceof protection are considered to be adequate, but not·providing the extra measure of capability that NRC
deems prudent. Licensee safeguards judged fair are
considered adequate to permit continued operation
only if the observed deficiencies do not pose an undue

Characterization of Malevolent Acts
Malevolent Acts

J
I

I

Sabotagea,bTheft, Diversion
ofSSNMa,c(Formula Quantities)

I

I

I

I

I

I

External Attack

Internal Attack

External Attack

Internal Attack

Violent External Attack
OR- Stealthful Attack
OR- Deceptive Actions(All by a Small Group)

An Individual(Including any Employee)OR
A Conspiracy-:- Violent External Attack
OR- Stealthful Attack
OR- Deceptive Actions(All by Several Persons)Insider Threat(Any Employee)

I

I

I

Attacker Characteristics

Attacker Characteristics

Attacker Characteristics- Well-Trained, Dedicated- Inside Assistance- Suitable Weapons
Hand-Carried Equipment- Well-Trained, Dedicated- Inside Assistance- Suitable Weapons- Hand·Carried Equipment- Operability as 2 or MoreTeams

Access to Detailed Knowledgeof Facilities- Items to Facilitate Theft

eln January 1979, the Commission directed the steff to conduct a study of the potential insider threat to the licensed nuclear industry. The Commission
expecu to receive the study report by early 1980.
bAs defined in 10 CFR 73.1 (8)(1\.
CAs defined in 10 CFR 73.1(a)(2).

Fuel Cycle Facilities
In January 1979, the NRC staff completed an
IS-month program of comprehensive evaluations ofsafeguards at the 11 licensed facilities which process, formula quantities of nuclear material. Four separate
field teams examined each facility with respect to:• Vulnerability to external assault.
• Physical security.
• Possible inside diversion paths.
• Nuclear material control and accounting.

risk to the public health and safety or common defense
and security during the short term required for theircorrection. Operations at facilities with safeguards
capabilities judged good or fair are permitted to continue only where agreement has been reached toundertake an immediate remedial program to bring
the licensee's safeguards program back to a state of ~high assurance. Continued operation of licensee
facilities where safeguards are judged poor will not be
permitted unless the observed deficiencies can becorrected immediately. To correct deficiencies, NRC

===================================================135
modifies licenses as needed to upgrade the systems,
and requires the applicable licensees to make thenecessary changes. NRC then conducts special followup inspections to ensure compliance.
Eight formal reports on findings and corrective actions concerning safeguards were sent to Congress during fiscal year 1979. The other three are completed
and are being sent to Congress early in fiscal year
1980.
Table 1 summarizes the teams' initial assessments of
the 11 fuel facilities examined. As noted above, if a
facility did not receive a high rating, it was required totake corrective actions designed to improve safeguards
to the desired level of high assurance. Licensees wereallowed varying amounts of time to correct deficien-,cies, depending on how long it would take for permanent corrective action; however, interim measures had
to be put into effect immediately. Inspectors from
NRC's Office of Inspection and Enforcement routinely
inspect both the interim and the permanent corrective
measures taken.
By the end of fiscal year 1979, all required corrective actions for the 11 facilities had been identified,and interim measures required by NRC had been put
into effect in those cases where permanent improvements had not been completed. As a result of
these improvements coming out of the comprehensive
review program, all facilities were judged to provide
high assurance of protecting against the hypothesized
threat. However, NFS Erwin experienced an excessiveat fuel cycle facilities during fiscal year 1979 included
more than 8,000 hours of on-site safeguards inspections at 14 fuel cycle facilities (those authorized topossess formula quantities of unirradiated SSNM in an
unsealed form). These inspections revealed 73 items of
noncompliance 'with safeguards requirements. (SeeTable 2 for summary of inspections.) The NRC took
particularly significant action (a shutdown order)
against one major licensee, Nuclear Fuel Services,
Inc., operator of a plant at Erwin, Tennessee, where
the inventory difference for the bimonthly physical inventory exceeded the upper limit specified in the conditions of the license. The licensee had also exceeded
the limit of error associated with the inventory difference. Based on inspection results, NRC has judged
the safeguards performance of the remaining licensees
as satisfactory.

Transportation Activities
Spent Fuel Shipments. In July 1979, NRC imposed
new safeguards requirements on licensed spent fuelshipments. A recent, government-sponsored study*
suggested that high-explosive sabotage of a spent fuelshipment, should it occur in an area of high population density, might produce serious radiological consequences. Although NRC does not have information indicating that a sabotage threat against spent fuelshipments exists in the United States, additional

Table 1. Original Field Evaluation Ratings, Before Corrections, at 11 Facilities
CRITERIA
EXTERNAL ASSAULT

a

Poor

Poor-Fair

Fair

RATINGS
Fair-Good

Good

Good-HighHigh b1a

The NRC staff issued an immediately effective order increasing the level of safeguards.

b Safeguards evaluation goal.

inventory difference and was shut down in late
September 1979, as discussed below.
On November 28, 1979, the Commission publishedthe Physical Protection Upgrade Rule, requiring
licensees to provide strengthened physical protection
for formula quantities of special nuclear material.
These new requirements will become effective March
25, 1980.
Inspection and Enforcement at Fuel Cycle
Facilities. NRC inspection and enforcement activities

safeguards were imposed as prudent, interim
measures, until a current research project confirms or
refutes the potential for such consequences.
The interim rule requires that licensed spent fuelshipments avoid, where possible, highly populated urbanized areas, and that the licensee obtain NRC approval of the proposed route before such shipments are• Ducharme, A.R., et al., "Transport of Radionuclides in Urban
Environs: Working Draft Assessment," SAND-77-1927, Sandia
Laboratories, May 1978.

136

Table 2. Safeguards Inspections During FY 1979; at Fuel Cycle Facilities

ar

Number oj
Items oj
Noncompliance

Number oj
Unannounced
Inspections

390 (13/377)1>
787 (108/679)

0
3 (0/3)83
77

5 (0/5)126 (0/126)

4 ( /4)100

12 (4/8)
5 (2/3)7 (4/3)5 (3/2)2 (1/ 1)25 (6/19)

468 (102/366)
388 (30/358)
539 (174/365)
308 (45/263)
65 (50/15)
1,644 (117/1,527)

6 (115)3 (2/1)6 (2/4)3 (3/0)4 (4/0)18 (2/16)

83
100
100
100
100
80

5 (3/2)289 (51/238)

4 (4/0)100

10 (317)
14 (3/11)444 (126/318)
1,266 (155/1,111)

2 (2/0)5 (0/5)100
86

13 (1/12)

1,110 (72/1,038)

12 (0/12)

46

Development Laboratory,
Cheswick, Pa.

12 (4/8)394 (117/277)

3 (0/3)75

TOTAL

140 (381102)

8,218 (1,16017,058)

73 (20/53)

83

Strategic Fuel Facilities
1. Babcock & Wilcox, Apollo, Pa.
2. Babcock & Wilcox, Leechburg, Pa.
3. Babcock & Wilcox, Lynchburg

Research Center, Lynchburg, Va."

Number
Sajeguar
Inspections

Number oj
Manhours oj
Inspection Onsite

12 (lI11)b13 (3/10)

4. Babcock & Wilcox, Naval Nuclear

Fuels Division, Lynchburg, Va.
5. Exxon Nuclear, Richland, Wash. e
6. General Atomic, San Diego, Cal.
7. General Electric, Vallecitos, Cal.
8. Kerr McGee Nuclear, Crescent, Okla.t'
9. Nuclear Fuel Services, Erwin Tenn.
10. Rockwell International, Canoga

Park, Cal.
11. Texas Instruments, North

Attleboro, Mass.
12. United Nuclear, Montville, Conn.
13. United Nuclear, Wood River
Junction, R. 1.
14. Westinghouse, Plutonium Fuela

Based on information on file as of November 5, 1979.

h For numbers in parentheses, the first number refers to physical security inspection activities; the second refers to material con-trol and accounting inspection activities.
t' These facilities are either not operating or not holding formula quantities in unirradiated form.

made. Data used to identify the applicable urbanizedareas are from the 1979 Supplemental Report of theu.s. Bureau of the Census. * For NRC route-approval
purposes, the following criteria define which
populated areas are to be avoided:
A. Any "urbanized area" with more than one
million persons.
B. Other "urbanized area" containing a city withmore than 500,000 persons.
C. The boundary of any city, other than that included in A and B (excluding the rural part of an
extended city), with more than 100,000 persons.
In addition, NRC's interim rule requires the following measures to be taken to guard against an attempt
to sabotage a spent fuel shipment:• Arrangements to call upon law enforcement agen-• u.s. Bureau of the Census, Supplemental Report, 1970 Census
oj Population, "Population and Land Area of Urbanized Areasfor the US: 1970 and 1960," Series PC (SI)-108, April 1979.

cies along the route for emergency assistance, if
needed.
• Mobile communications equipment accompanying each shipment in transit.
• Status report, every two hours, from a shipmentin transit to a central location.
• Capability to immobilize the transports, should ahijack be attempted.
• At least two individuals with each shipment,
highly trained in emergency response procedures.
• No intermediate halts, where practicable, except
for necessary refueling and food stops, and continuous occupation of the vehicle during
necessary halts.
• Procedures for coping with threats and emergencies during transit, including arrangements forcalling local law enforcement assistance.
• Armed escort, or local law enforcement agency

137

escort, if transport is necessary (and NRC·
approved) through a highly-populated area.
SSNM Shipments. During 1979, NRC requirements
for safeguarding shipments of formula quantities ofhighly enriched uranium or plutonium remained thesame as those during 1978.
Route Surveys. NRC Safeguards teams conduct fieldsurveys over transportation routes proposed for spent
fuel or strategic special nuclear material shipments.
These surveys obtain information for NRC contingency planning and route approval considerations. During such surveys, the teams coordinate with local law
enforcement agencies along the way to increase theirawareness and knowledge of the shipments and to
identify local contacts that can be helpful, if needed.

A shipment of foreign irradiated fuel is transferred from ship toflatbed trailer for sliipment to the DOE's Savannah River Plant
near Barnwell, South Carolina. Under the Atoms for Peace agreement the U.S. sells nuclear fuel to foreign countries with the condition that spent fuel will be returned to the U.S. for storage.

Rail cars can carry several truckloads of fuel elements. The picture
shows several spent fuel casks with special covers now mounted ona rail flatbed car.

During fiscal year 1979, the NRC teams surveyed
five routes for shipments of formula quantities of
special nuclear material and four routes for shipmentsof spent fuel. They collected data in the field, traveling approximately 6,400 road miles through 33 States,
and meeting with approximately 270 local and Statelaw enforcement agency representatives along theway. Licensees transporting nuclear materials received route profile reports which described appropriate
law enforcement contacts and communications along
the way.
During the year, NRC determined the adequacy ofsafeguards by both the licensing process and the inspection of all licensed shipments involving special
nuclear material. These inspections covered all
domestic shipments and the domestic segments of the

A special steel cover is placed over the cask and bolted to the bedof the flatbed trailer. This reduces visibilitY of and accessibility to
the cask while in transit to a railhead.

Regulations require the placement of buffer cars between any car
carrying special nuclear material and other cars in the train bothto provide safety space between the shipment and the other cars
and to allow a 360 0 view of the shipment itself. This photo shows
shipment as observed from the observation cupola of the caboose,looking forward towards engine. The shipment was observed fromthe rear by the conductor and from the engine by the train masterand/or engineer throughout.

138

import and export shipments, including all storage
and transfer points. Twenty-seven shipments werecovered, involving 689 man-hours of inspection activity. No items of noncompliance were noted.

Reactor Safeguards
The main factors in evaluating the adequacy ofsafeguards at power reactor facilities are:• Licensing Reviews (including field reviews ofequipment and operations) of licensee security
plans prepared in response to the requirements to
10 CFR 73.55. ("Requirements for Physical Protection of Licensed Activities in Nuclear Power
Reactors against Industrial Sabotage.")• Inspection for licensee compliance with the approved security plan.
The NRC staff is also planning a future program toevaluate the practical effectiveness of safeguards as
implemented at licensed operating power reactors
throughout the U.S.

rule" procedures in vital areas, and additional physicalcompartmentalization within such areas. (Two-man
rule procedures are those where two employees
observe each other's activities, in order to minimize
the opportunity for malevolent acts by an insider.)The Commission also is considering various programsfor determining trustworthiness of reactor facility
employees authorized to enter vital areas.
There have been some construction and equipment
delivery delays in the implementation of certain
safeguards measures. Therefore, certain facilities areusing approved interim measures, such as additionalarmed guards, until final systems components can beinstalled and their operation verified.
As of February 23, 1979, new security requirements
for all power reactors went into effect. Since that time,
69 power reactors have been inspected to determine
compliance with the new security requirements. With
the exception of escalated enforcement actions,discussed in Chapter 7, the concerned licensees dealt
promptly with all items of noncompliance.
Status of Safeguards at Non-Power Reactors. All
licensed non-power reactors have operative security
plans, required by 10 CFR 73.40, ("Physical Protection: General Requirements at Fixed Sites") for protection against sabotage. Specific safeguards measures for
non-power reactors with less than formula quantities
of SSNM are not defined in NRC regulations (although
the new NRC regulations of July 1979 for these quantities of material apply to any facilities possessing suchmaterials). However, those in effect at such reactor
sites include:• Identification of security areas.
• Security organization .
• Access controls.

NRC safeguards regulations require that all personnel entering
critical areas of nuclear reactor facilities be thoroughly searched asa measure of assurance against possible installation sabotage. Here,NRC Resident Reactor Inspector Leif J. Norrholm is subjected to a
pat-down search for weapons or possible sabotage devices at theSalem 1 Nuclear Generating Station in New Jersey.

Status of Safeguards at Power Reactors. All
operating power reactor licensees have put into effectapproved physical security plans meeting the generaland specific requirements of 10 CFR 73.55, with some
aspects of measures employed against the inside threat
being further refined. In this connection, the Commission has deferred the required implementation of certain defensive measures against potential sabotage by
a licensee employee inside the facility pending further
evaluation of the need and alternatives. These defensive measures include pat-down searches, "two-man

Shipments of seecial nuclear material are guarded during allphases of tranSit. Here, guards prepare to offload a truck shipment
at a user facility. NRC monitors and inspects SNM shipments,
surveys shipping routes, supervises guard and driver training, and,in cooperation with the Department of Transportation, develops
standards for every aspect of nuclear materiar transportation.

===================================================139• Methods for detecting and reporting intrusions.
Also, non-power facilities with formula quantities
of SSNM that are not self-protecting have activatedanti-theft measures meeting the specific requirementsof 10 CFR 73.50 ("Requirements for Physical Protection of Licensed Activities") and 10 CFR 73.60 ("Additional Requirements for the Physical Protection ofSpecial Nuclear Material at Non-Power Reactors").

vestigation which formalized procedures for information exchange and coordinated response actions between the two agencies. NRC is now coordinating
three similar interagency agreements with the Federal
Aviation Administration, the National Security Agency, and the Bureau of Alcohol. Tobacco and Firearms.

Inspection and Enforcement at Reactors. NRC inspection and enforcement activities at reactors also
provided a measure for judging the effectiveness of
safeguards. During fiscal year 1979, the NRC expended nearly 8,000 hours in on-site safeguards inspectionsat power reactors, and 2,300 hours at non-power reactors and research facilities. These inspections revealed
385 items uf noncompliance with safeguards requirements (see Table 3). NRC has issued two Immediate
Action Letters that identify measures which licensees
must take to improve their safeguards systems. Threecivil penalties, totaling more than $25,000, were
issued against three licensees.

In December 1978, a licensee shipped four drums
containing 4.5 kilograms of highly enriched uranium
to Romania, with each drum sealed by the NRC. Upon
arrival at the port of embarkation in New York, allfour seals were found broken. The NRC inspector examining the drums and seals decided that the contentsof the drum had not been disturbed. Consequently, heresealed the containers-without inspecting the contents-and permitted the export of the shipment.
When the containers arrived in Romania,. the seals
that were affixed in New York were found to be intact.
The IAEA, upon request, examined the shipment upon
arrival, and confirmed that the nuclear material con-

SAFEGUARDS EVENTSFISCAL YEAR 1979

Table 3. Reactor Safeguards Inspections During FY 1979 adt

Number
Safeguar
Inspections

Manhours of
OnsiteInspection

Number of
Items of
Noncompliance

Percent of
Unannounced
Inspections

Power Reactors

190 (163/27)b

7,778 (7,010/768)b

340 (330/1O)b96

Non-power Reactors

55 (41/14)

926 (743/183)

19 (17/2)

96

Research & Specialty
Reactors

53 (5148)

1,399 (152/1,247)

26 (9/17)

76

Facilitya Based on information on file as of November 5, 1979.
b For numbers in parentheses, the first number refers to physical security inspection activities, the second refers to material con-trol and accounting inspection activities.

Contingency Planning
During fiscal year 1979, the NRC staff reviewed andapproved safeguards contingency plans developed byeach of the 19 fuel cycle facility licensees authorized tohave formula quantities of strategic nuclear material.
During this time, the staff also reviewed 28 of 61
safeguards contingency plans developed by powerreactor licensees and applicants. Each of the powerreactor contingency plans evaluated required certain
modifications or additions. None was given final approval during fiscal year 1979.
At the national level, NRC concluded a memorandum of understanding with the Federal Bureau of In-

tents were as shipped by the licensee. As a result of thisevent, NRC has changed its procedures for accounting
for the seals that allow inspectors at the destination todetermine whether anyone has replaced or tamperedwith them.
In January 1979, the General Electric Company
reported to the NRC that an alleged theft of lowenriched uranium oxide, and an attempted extortion
had occurred at its fuel plant at Wilmington, North
Carolina. An individual was arrested on criminal
charges, and the material was recovered. The material
was of no safeguards significance because of theamount and enrichment involved. It represented aminimal health hazard, being less hazardous than

140

many industrial chemic~ls. As a .result of the incid~~t,the licensee reevaluated Its securIty and accountabIlIty
system for such materials. For some time, the N~C has
been reexamining requirements for the protectIOn of
nuclear materials of this type. As a result, new rules
are being proposed to prevent future incidents of t~is
type from occurring. (See Chapter 7 for further detaIls
on this incident.)
In May 1979, 62 of 64 new fuel assemblies at the
Surry Nuclear Power Station i~ Virginia w~re fo~nd tohave been coated with sodIUm hydroxIde, m anapparent attempt at sabotage. Subsequent investiga·
tions revealed that two licensee employees were
responsible for the act. They were later convicted a~dsentenced to prison. (See Chapter 7 for further detaIls
on this incident.) The NRC has subsequently taken
steps to modify its regulations to tighte~ access controls at such facilities.· Requirements bemg proposed
are contained in an Inspection and Enforcement
Bulletin, IE Bulletin No. 7416, "Vital Area AccessControls. "As a result of a physical inventory taken in August
1979 Nuclear Fuel Services (NFS) reported that theinve~tory difference at the NFS plant at Erwin, Tennessee, was in excess of the upper limit specified in the.
license. (An inventory difference occurs when the total
SNM listed in the account books from previous
measurements does not agree with the total SNM
measured in the most recent physical inventory.) The
licensee was ordered to shut down the facility andbegin an immediat~ i~vestigation and .re-inventory.
NRC sent an investigation team to the SIte to observe
the inventory and verify measurements being made.
While the re-inventory results partially explained theinventory difference, the discrepancy was not reduced
to a level normally expected as a result of uncertainties
in the measurement of nuclear materials. At the end of
fiscal year 1979, the facility remained shut d.own ~en­ding resolution of the problem.'" The NRC mvestIgation which included FBI and Department of Energy
participation, had not discovered any inf~rmation(other than the presence of the inventory difference
which is of itself indeterminate because of measurement uncertainties) to indicate that a theft of material
had occurred. However, the investigation had notbeen able to rule out that possibility.

SAFEGUARDS REGULATORYACTIVITIESFiscal year 1979 was an active period in the development and adoption of NRC regulations designed to improve nuclear safeguards. Attempting to solve several. . On January 21,1980, the facility was permitted to return to production operations after implementation of significant improvements in the physical security, internal control, andmaterial accounting systems.

major safeguards problems also constituted an important part of NRC's activities. Solutions were notalways found and efforts to solve some problems must
continue into 1980 and beyond, if necessary.
Physical Security. In November 1979, the Com~is­sion published the final version of a new regulatIOn
designed to upgrade the physical protection of formulaquantities of special nuclear material. Any facility
holding or transporting five formula kilograms of suchmaterial is subject to the new rules. The only temporary exception is non-power reactors, for the reasonsindicated below. The regulation will become effectivein March 1980. Licensees are expected to implement itby the fall of 1981. The rule indicates performancestandards to be met by licensees and presents specificstatements about the kinds of threats, from insiders
and outsiders, that their safeguards should be able towithstand.
Most non-power reactors are research reactors
operated by universities. Applying the new rules musttake into account the unique characteristics of auniversity and its reactor. The universities are concerned that applying the strict new physical security
requirements of the regulation will force them to shut
down their reactors. They cite unacceptable costs andimpact on their education programs. They also cite
specific design and fuel features of their reactors. Someof these views appear to have merit, and the NRC isconsidering whether a balance can be struck between
the specific new requirements and alternative ways t?
achieve the necessary safeguards performance. ThIS
problem was not resolved in 1979, and continues to bestudied.
Another important problem to be resolved in 1980
relates to the regulation upgrading power reactor
physical security safeguards. When first issued in 1977,
the rule called for conducting either a physical or an
instrument search for the detection of prohibitedmaterial. The Commission was petitioned to eliminate
the possible interpretation of "physical search" as requiring a "pat-down" search. The issue is under study,
and the staff is considering alternatives for the mosteffective search techniques to be employed. In the interim instrument searches are routinely employed,
suppl~mented by physical searches when circumstances dictate.
In July 1979, final regulations were published toprovide new physical protection requirements for
special nuclear materials in less than five formulakilogram quantities-materials of moderate- and lowstrategic significance (Category II and Category III
materials). These regulations make the U.S. rules consistent with International Atomic Energy· (IAEA)
worldwide standards.
Another new rule issued in fiscal year 1979 placed
carriers of five or more formula kilograms of special
nuclear material under NRC regulation. Formerly,
the shippers and receivers, but not the actual

141

transporters of material, had been under NRC license.
A general license has been issued to the carriers, making them directly subject to NRC requirements and inspections.
Requirements of the new interim rule for spent fuelshipments have been discussed in detail earlier in thischapter. The staff must now reassess the need forchanges based on public comments, and as may be required by the result of recent research. Among theissues being addressed are routing restrictions,
preemption of local ordinances, call-in procedures,and handling of vessels carrying spent fuel.
Transient shipments of formula quantities of special
nuclear material are also a matter of concern. A transient shipment is one that temporarily uses U.S.
facilities while moving from one foreign country toanother. NRC is preparing a regulation that would
require protection of such shipments, which usually
are carried by an aircraft transiting a U.S. airport.
Enforcing such a regulation may pose problems,
because it would require that NRC obtain advancedetails about the shipment, sometimes from sources
outside NRC's immediate control.
Transient shipments of spent fuel are a matter ofmore recent concern. The staff has taken note of growing public concern about the need to provide
safeguards protection for such shipments. This concern was highlighted when Representative Heftel and
the Governor of Hawaii urgently requested that the
NRC adopt regulations to protect transient shipmentsof spent fuel. These requests came as a result of an
unscheduled refueling stop, at Honolulu, by afreighter carrying spent fuel. The NRC plans toanalyze the alternatives involved in prOViding suchprotection and possible regulatory changes to implement such protection.
Along with its concern about shipments of spent
fuel, the NRC also is trying to estimate the potential
hazards of sabotage (or theft, if that should occur) at
high level nuclear waste storage sites. Conceivably,
the radioactive dispersal hazards might be similar to
those resulting from sabotage of spent fuel. The staff isalso analyzing the alternatives involved in transporting Three Mile Island wastes to disposal sites. Whenthe results of these analyses are known, the staff willaddress the issue of what safeguards measures, if any,
should be required for nuclear waste activities.
The Kemeny Commission's report on the Three Mile
,Island accident indicated that human attitudes and'practices were a principal contributory fador to theaccident. In the area of nuclear security, peopleparticularly security managers and guards-play avital role in ensuring that safeguards systems achieve
their intended purpose at all licensed facilities. In
response to these concerns, the NRC staff has begun toreview how well people perform in the safeguards
area. The staff will make recommendations to improve such performance, if appropriate.

The NRC role in safeguards training includes setting training requirements for transport security personnel, ensuring appropriate
cooperation and communications between shipment security
guards and local law enforcement agencies, and monitoring suchactivities through inspections and reports. Shown here are two examples of training inspected by NRC in 1979. The photos show
[from top] a formal security-officer weapons training class, andscenes from a staged accident involving a "nuclear shipment" inwhich accident response measures were practiced.

142

Material Control and Accounting. While not always
immediate or inexpensive, solutions to security problems at NRC licensed facilities seem to be more
obvious and straightforward than solutions toaccountability problems. An example of the intractability of these problems was provided during fiscal
year 1979 by the occurrence of a large inventory difference at the NFS fuel plant at Erwin, Tenn., as
described earlier in this chapter. (An inventory difference occurs when the total SNM listed in theaccount books from previous measurements does notagree with the total SNM measured in the most recent
physical inventory.) Another example of material control and accounting problems is provided by
NUREG-0627, an NRC report on the Nuclear Materialand Equipment Corporation (NUMEC) situation in
the mid-1960·s. (The report was requested by Congressman Udall and prepared by NRC staff during
fiscal year 1979.)
For years NRC (and previously the AEC) has usedinventory differences to signal accountability problems or out-of-control processing situations. However,inventory differences are based on periodic plant-wide
inventories and are not a timely indicator of a loss ofmaterial. Moreover, the causes of unusual or excessive
inventory differences are not always clear, even after
extensive investigation. As a result, the NRC staff is examining several alternatives to relying on inventory
difference as a primary indicator of accounting problems.
Currently, the staff is formulating a major rule
aimed at improving the level of safeguards assurance
provided by material control and accounting systems.
The goals are to provide more timely material control
and accounting indicators which can be resolved moreclearly, and which will better locate the accountability problems within a licensee's plant. A draft rule
is planned by the end of fiscal year 1980.
NRC/IAEA Interaction. In 1979, NRC staff accelerated preparations for the application of lAEA
safeguaras to U.S. nuclear facilities other than thosehaving direct national security significance. The
US/IAEA Safeguards Agreement has been approved by
the lAEA. The President has submitted it to the Senate
for ratification as a treaty. The Agreement will enter
into force when the U.S. notifies the IAEA that its constitutional and statutory requirements have been met.
(See Chapter 9 for detailed discussion of internationalsafeguards.)

SAFEGUARDS RESEARCH ANDTECHNICAL ASSISTANCEThe NRC safeguards contractual program includes
both research (long-term, comprehensive effort) andtechnical assistance (short-term efforts supporting
operational assignments). In fiscal year 1979, about$12 million was spent on safeguards research andtechnical assistance. Approximately $5 million of thetotal was spent on research projects and the remaining$7 million on technical assistance projects. The Commission reviewed and approved all safeguards contracts exceeding $20,000 in funding, as required by theCongress.
The safeguards research program has now
developed to the point where the products are being
tested in applications to assist (a) the formulation ofregulations; (b) the determination of safeguards adequacy; and (c) the assessment of the effectiveness oflicensee safeguards systems.
Fiscal year 1979 research projects that have helpedor are expected to improve safeguards programs are
the following:• "Effectiveness Evaluation Methods for PhysicalProtection of SNM in Transit" Proiect. Resultsfrom this program have helped evaluate guard
levels required for an in-transit physical protection system established in the physical protection
upgrade rule.
• The "Effectiveness Evaluation Methods forMaterial Control and Accounting" research profeet, which has provided technical inputs todevelopment of the proposed material control andaccounting upgrade rule, and associatedguidance.
• The "Insider Crime Analogous to the Potential
Threat to Nuclear Programs" study, which is expected to aid in the formulation of prudent standards and regulations related to the potential insider threat.
• The "Spent Fuel Cask Vulnerability Program,"
results of which will help formulate policy on
safeguarding shipments of spent fuel from lightwater reactors. They will help confirm and/or
modify, as necessary, safeguards regulations protecting spent fuel shipments.
• The "Nuclear Power Plant Design Concepts for
Sabotage Protection" proiect. Design alternatives
and damage control measures for nuclear power
plants were studied in order to improve their inherent protection against sabotage. A recent
typical plant design * was selected andcharacterized to provide a baseline against which
the effectiveness and impact of proposed changes
will be measured. A set of potentially usefuldesign alterations, as well as methods to mitigate
damage, were identified. These features havebeen reviewed and the most promising alter-·The Standardized Nuclear Unit Power Plant System is a system
identified by Bechtel Corporation. It takes advantage of standardized
engineering and installation practices for the purpose of simplifying
licensing and acceptance reviews.

143

natives will be selected for more comprehensive
evaluation.
Fiscal year 1979 research projects directed toward
improving safeguards adequacy and effectiveness include:• The "Effectiveness Evaluation Models for Fixed
Site Physical Protection" project. The method
developed in this project identified vital areas
within reactor facilities, using computerized,
generic fault-tree techniques. It was applied tomore than 27 pressurized water reactor and boiling water reactor sites, and has proved an effective aid in licensing and evaluating power reactors. As a result of last year's feasibility test of the
Safeguards Automated Facility Evaluation
method, some Safeguards Automated Facility
Evaluation modules were further developed andapplied, on a trial basis, to several reactor

facilities. This method will be used extensively infiscal year 1980, with the vital area analysis
already discussed, to help evaluate the adequacyof safeguards at power reactor facilities. The
Safeguards Network Analysis Procedure is anevaluation tool to help in physical security field
evaluations of operating facilities and to evaluate
proposed licensee safeguards. This procedure may
also be useful for formulating insights into guard
tactics and strategies planned as responses topossible terrorist attack.
• The "Effectiveness Evaluation Methods for Material Control and Accounting" project. The structured assessment analysis method was substantiallydeveloped in fiscal year 1979 for use in assessing
the effectiveness of material control and accounting safeguards at fuel-cycle facilities. The structured assessment analysis assists in analyzing thevulnerability of a facility to both insider and

The improvement of safeguards techniques often means combining
existing methods or equipment toward greater precision or speed in
obtaining results. Dr. Warren McGonnagle of NRC's Region III
Office of Chicago is shown here observing measurements during a
non-destructive enrichment assay of nuclear fuel material. The
material, physically located in a detector (under McGonnagle's
hand), is automatically evaluated in the multi-channel analyzer on

his ri2ht. Calculations then .can either be printed by the attachedrecoraer (left of the analyzer) or displayed on the computer at theright of the photo. The canister beneath the detector feeds nitrogen
into the deVice to cool the test sample. Data from these independent NRC assays are compared with information supplied by commercial firms for verification of enrichment percentages or otherproperties.

144==================================================
outsider adversaries with authorized andunauthorized access and extensive capabilities.
The method is scheduled for testing in fiscal year
1980.
• The "Communicated Threat Credibility Project. "This project provides multidisciplinary tools forinvestigating the credibility of communicatedthreats and for providing advice to the Department of Energy, the NRC, the Federal Bureau of
Investigation, and other appropriate agencies
during an actual or perceived emergency from
nuclear extortion threats.
Technical assistance projects were conducted by the
major program offices to support their operational
missions. These projects ranged from helping establish
a technical basis for determining safeguards requirements for byproduct materials, to providing
assistance in developing NRC's physical security
upgrade rule. Technical assistance projects of greaterthan $20,000 were approved by the Commission, as
required by Congress.

SAFEGUARDS MANAGEMENT
Under the NRC's new lead office management concept, the Office of Nuclear Material Safety and
Safeguards (NMSS) is the lead office for safeguards
and is responsible for integrating and coordinating theoverall NRC safeguards program.
Safeguards Consolidation. In August 1979, NRC
decided to transfer all reactor safeguards functions
formerly in the Office of Nuclear Reactor Regulation
to NMSS, which also has primary safeguards
regulatory responsibility for fuel cycle facilities andtransportation activities. The consolidation was effective on October 1, 1979. Staff members in NMSS coordinate their reactor safeguards licensing activities with
the Office of Nuclear Reactor Regulation. This coordination will involve administrative maintenance ofthe safeguards portion of reactor licenses. It will also
involve areas of concern common to both reactor
security requirements and reactor safety matters.
Reactor safeguards functions transferred to NMSS
include the following:• Safeguards licensing reviews for power and nonpower reactors.
• Generic physical security policy and guidancedevelopment for reactors.
• Work on the potential implementation of theUnited States/International Atomic Energy Agency Safeguards Agreement.
• Administration of reactor safeguards technicalassistance contracts.
The Office of Inspection and Enforcement will continue safeguards inspection functions. The Office ofStandards Development will continue to develop
safeguards standards. The Office of Nuclear Regulatory Research will continue both its present and future
safeguards research function.
Integrated Program Plan and Safeguards Technical
Assistance and Research Coordinating Group. Each of
the major program offices participates in the planningand implementing of NRC's domestic safeguards program. The Safeguards Technical Assistance andResearch Coordinating Group provides inter-office
coordination of NRC-contracted safeguards activities.
An Integrated Safeguards Program Plan wasdeveloped and sent to the Commission in January
1979. This program plan, which will help coordinate
safeguards activities in NRC, will be updated in fiscal
year 1980 after the Commission issues its policy, planning, and programming guidance. NRC's Budget
Review Group, the Executive Director for Operations,and the Commission also review and approve the entire NRC safeguards program during the annual NRC
budget review process.

6
Waste
Management

Waste shipments are clearly identified by
NRC-approved placards.

The NRC waste management function was elevatedto divisional status in 1979 under the NRC Office of
Nuclear Material Safety and Safeguards (NMSS). The
new Division of Waste Management consists of fivebranches which carry out a number of functions thatwere formerly among those of the Division of Fuel
Cycle and Material Safety:• The High-Level Waste Technical Development
Branch-responsible for high-level waste
regulatory development and development of thetechnical bases for high-level waste licensing andregulation.
• The High Level Waste Licensing Management
Branch-responsible for licensing high-level
waste commercial repositories.
• The Low-Level Waste Licensing Branchresponsible for low-level waste licensing andregulation.
• The Uranium Recovery Licensing Branchresponsible for licensing and regulation of
uranium mills, heap-leach operations, commercial scale solution mining operations, andresearch and development (R&D) uranium
recovery operations. These types of operations
represent the first step of the nuclear fuel cycle.
Since large amounts of waste are generated as aresult of these operations, especially uranium
milling, it was decided that these operations
should come under the Division of Waste
Management.
• The Licensing Process and Integration
Branch- responsible for coordinating and integrating the entire NRC waste management program. In order to do this, the branch works withelements within NMSS, with other NRC offices,and with other governmental agencies having
waste management responsibilities, to ensure thatthe entire NRC program is well focused and proceeding on established schedules.

Overview of 1979 Activity
The main focus of NRC effort in 1979 for the highlevel waste program was in regulatory development.
The NRC is developing a comprehensive regulation for
high-level waste repositories-to be Part 60 of the
NRC regulatory code-in two parts, procedural andtechnical. The procedural part was published as a proposed rule for comment in December 1979. The technical part is expected to be published in early 1980
pursuant to an advance notice of proposed rulemaking.
the main focus of NRC work in 1979 for the lowlevel waste program has also been in regulatory
development. The NRC is developing a comprehensive
regulation for low-level waste disposal. This regulation will be Part 61 of the code. A preliminary draft of
the regulation has been completed and sent to variousorganizations for review. The draft will be 'madeavailable to the public in 1980.
A large part of NRC effort under the uranium
recovery program has been concerned with the licensing of uranium recovery facilities, and a significant
number of licenses were issued, renewed, andamended. In addition, a draft regulation for uranium
mills (Amendment to Part 40) was issued for publiccomment in August 1979. The supporting generic environmental impact statement (GElS) on uranium
milling was issued for public comment in April 1979.
A number of notable events in nuclear waste
disposal took place in 1979. There were only three
low-level waste disposal sites in operation at the beginning of the year, all of them located in Agreement
States. Two of the sites closed and then reopened, anda curtailment was placed on the amount of waste that
could be received at the third site. These actions further demonstrated the large regional imbalance in
low-level waste disposal locations and induced a146Interagency Review Group. As reported in the 1978
NRC Annual Report (pp. 93 and 94), the NRC staff
participated in the IRG study on Nuclear Waste
Management. (Because of NRC's status as an independent regulatory agency, the agency participated as a
non-voting member.) The IRG draft report was issued
in 1978 and the final report was issued in 1979. Manyof its recommendations affect the NRC, which hasreviewed the impact of these recommendations on its
program in 1979 and will continue to do so in 1980.

On July 16, 1979, a tailings dam near Grants, N.M., gave way,
releasing nearly 100 million gallons of radioactive water and sediment into the Rio Puerco. Flow from the break reached into
Arizona, some 75 miles down river. The break occurred as efforts
were being made to reinforce the dam, and heavy equipment onsite for that purpose enabled workers to stop the flow in a few
hours.

number of States to seriously consider the desirabilityof regional burial sites. Also, on July 16, 1979, a tailings impoundment failure occurred at the United
Nuclear Corporation uranium milling operation at
Church Rock, N.M. (New Mexico is also an Agreement
State.) A major effort was undertaken by the NRC toassist the State in correcting the situation. (See discussion under "Technical Assistance to Agreement
States," later in this chapter.)
It is important to note three studies which have affected and will affect the course of the NRC waste
management program. These studies are the Interagency Review Group (IRG) Study on Nuclear Waste
Management, the Congressionally requested NRC
Study on Regulation of Federal Radioactive Waste Activities; and the Congressionally requested NRC study
on Means for Improving State Participation in the
Siting, Licensing and Development of Federal Nuclear
Waste Facilities. Also of potential importance to the
NRC waste management program is the "confidence
hearing" on radioactive waste disposal to be held by
the Commission in 1980.

Federal Radioactive Waste Study. The NRC's
Authorization Act for fiscal year 1979 (P.L. 95-601) required the NRC to prepare a study on the regulation ofFederal radioactive waste activities. The study wascompleted in 1979 and it was issued as NUREG- 0527,
entitled "Regulation of Federal Radioactive Waste Activities." Two principal recommendations came out of
the study. The first was that NRC licensing authority
should be extended to cover all new Department ofEnergy (DOE) facilities for disposal of transuranic
waste and non-defense low-level waste. This recommendation was consistent with one of the IRG recommendations. The second was that a pilot program
should be established to test the feasibility of extending
NRC regulatory authority on a consultative basis to
DOE waste management activities not now covered by
NRC's licensing authority, or to the new facilities citedin the first recommendation. The pilot program would
focus on a few specific DOE waste management activities and would result in a report to Congress on thefeasibility of an NRC consultative role in existing DOE
waste disposal and storage activities. The decision onwhether to extend NRC regulatory authority and toestablish the pilot program and on what waste
management activities the program should include
was considered one for the Congress to make. If theCongress decides that the NRC should implement
these recommendations, it will significantly affect
NRC's current and future waste management programs. The exact impact cannot be assessed until
specific legislation is proposed and implemented.
Improving State Participation. The NRC's
Authorization Bill for fiscal year 1979 (P.L. 95-601)
also required the NRC to prepare a study on means for
improving the opportunities for State participation in
the process for siting, licensing, and developing
nuclear waste storage or disposal facilities. The study
was completed in 1979 and it was issued as NUREG
0539, entitled "Means for Improving State Participation in the Siting, Licensing, and Development ofFederal Nuclear Waste Facilities."
There were a number of recommendations as aresult of the study. The Commission recommended theestablishment of a planning council consisting ofFederal and State representatives, to be supported by asmall administrative staff and Federally financed. A
review capability should be established under the==================================================147
direction of the planning council in order to enable theStates to make technical evaluations of waste management technology and Federal waste management activities. The review capability should also be Federally
funded. These recommendations were consistent with
the IRG recommendations. In addition, the Commission recommended that measures be taken to involve
the States throughout the process for planning, siting,developing, and licensing nuclear waste storage
disposal facilities. It is also recommended that theCongress establish a grant program to allow the States
to partiCipate more fully in the Federal Waste
Management program. Federal agencies should consider such transportation related issues as shipping
routes, emergency planning, enroute liability, shipping containers, and the like, in their overall waste
management activities and should develop institutional arrangements as appropriate for consulting withsion's degree of confidence that radioactive wastes produced by nuclear facilities will be safely disposed of,and to determine when any such disposal will beavailable, and whether such wastes can be safely
stored until they are disposed of. Notice of the proceeding appeared in the Federal Register in October
1979, and the hearing will take place in 1980 and
1981. The proceeding has been initiated in response to
the decision of the U. S. Court of Appeals for theDistrict of Columbia Circuit in State of Minnesota v.
NRC, 602 F.2d 412,but is also a continuation of proceedings previously conducted by the Commission inthis area. The notice described the procedures theCommission will employ and how members of the
public can participate. The results of the hearing andany rules issuing therefrom may have an effect on
NRC's current and future waste management program. (See also "Commission Decisions," in Chapter
13.)
The three sections which follow describe the 1979
accomplishments of the NRC waste management programs dealing with high-level waste, low-level waste,
and uranium recovery. Each section discusses nearterm objectives of the program and activity during thereport period in regulatory development, licensing,and associated matters.

HIGH-LEVEL WASTE MANAGEMENT
Regulatory Development

NRC continued to study ways to improve FederallState cooperation in waste storage matters in 1979, as visits to Agreement-State
activities were stepped up. Representatives of several NRC program offices are shown liere during a briefing on low-level waste
storage monitoring techniques by officials of the Barnwell, S.C.,
storage site and South Carolina State offices.

the States in a timely manner. Lastly, the Commission
recommended that legislation for improving State participation in the Federal Waste Management Program
should provide recognition of the legitimate concernsof host States; considerations affecting a State concurrence or veto, if authorized by law, were identified.
If the Congress elects for the NRC and other Federal
agencies to implement any or all of these recommendations, these actions will affect NRC's current andfuture waste management programs. The exact impact
cannot be assessed until specific legislation is proposed
and implemented.
Confidence Hearing. The NRC decided in 1979 toconduct a generic proceeding to reassess the Commis-

NRC continued its high-level waste regulatory
development effort in 1979 with the objective ofdeveloping and publishing a draft regulation (10 CFRPart 60) and supporting environmental impact statement (EIS). The regulation as currently envisioned
will be published in two parts: the procedural requirements and the technical requirements. The procedural
portion would contain sub-parts covering general provisions, licenses, and participation by State governments. The technical portion would contain sub-parts
covering performance objectives and technicalcriteria, physical protection, quality assurance, andemergency plans. Particular emphasis is being placedon waste form performance requirements and geologic
site characterization issues. In December 1979, theprocedural portion of the regulation was published asa proposed rule for public comment. The technicalportion of the rule is expected to be published in early
1980 pursuant to an advanced notice of proposed
rulemaking. Work is also continuing on a supporting
environmental impact statement which would be
published with the proposed technical rule in 1980.
Work began in 1979 in developing regulatory guides
to support the regulation. These include format andcontent guides for the safety analysis report, the environmental report, and reports detailing DOE plans
for site characterization work. These guides will be
published for public comment in 1980.

148

Additional regulatory guidance will he provided to
DOE in the form of technical directives. The technicaldirectives that were under development in 1979 andwhich will be issued in 1980 will cover generic topics
addressing site selection and characterization,
repository design, and waste form. Work also con~tinued in 1979 on identifying research needs.
In 1979, work was begun on outlining license review
procedures both to aid the staff in establishingpriorities for research and regulatory gUides and toprovide DOE with guidance on how NRC will con~duct its review.

Licensing
NRC continued its efforts in 1979 to develop acapability to review a license application for a highlevel waste repository. The development of models forassessing radionuclide transport in bedded salt wascontinued and is expected to be completed in 1980. A
model for assessing the safety and environmental risksof a repository after sealing was delivered to the NRC
by the contractor so that NRC could test and evaluate
the model.

Assessment of DOE High-Level
Waste Management Program
The NRC has continued its assessment of the DOE
high-level waste management program in 1979. The
NRC reviewed and provided comments to DOE on thedraft EIS for the proposed Waste Isolation Pilot Plant(WIPP) and the draft GElS on the management ofcommercially generated radioactive waste,
The NRC initiated in 1979 a program to criticallyassess the DOE high level waste management proCOMPUTER

ELECTRON
LINEARACCELERATORBARRELSCANNERMECHANISM

llNAC
CONTROLCONSOLEBREMSSTRAHLUNGBEAM COLLIMATORCYLI NDR ICAl
WATER TANKOne of NRC's continuing objectives is the improvement of nuclear
material inventory and accounting techniques. This cutaway drawing shows the main features of a mobile measurement system used
to identify, measure and record uranium and plutonium contents

ot metal waste drums. The barrel ,scanner at rear (left) of themobil unit remotely places, lifts and "reads" the container, andtransmits readings throu2h, analytical devices to the recording instruments in the mannecf area of the trailer.

===================================================149
gram. DOE and its contractors have made formal
presentations to NRC on various phases of the DOE
program. On November 15-16, a meeting was heldwith the Office of Nuclear Waste Isolation and other
DOE contractors at Columbus, Ohio, to formally inaugurate NRC's assessment program. Arrangements
are being made to maintain an overview of all DOE
activities in high-level waste management bysystematically receiving and reviewing all documents
generated by the DOE program. Task groups havebeen established to perform an initial, limited assessment of DOE activities in waste packaging, repository
siting, and repository design. Comparisons will bemade between needs identified in NRC's draft regulation and information expected to be generated by
DOE programs. Finally, plans have been prepared for
conducting a comprehensive critical assessment of the
DOE repository siting and in-situ testing programs.

waste classification contractor is characterizing
wastes, waste forms, and waste sources in addition torecommending requirements for safely disposing thewaste. The volume reduction contractor is investigating various volume reduction techniques including compaction and incineration. The contractor
is also performing economic analysis for the various
techniques.
Other contractual efforts are planned to develop
specific technical criteria for disposal of wastes inmined cavities and engineered structures, and to investigate in detail requirements for disposal of waste
generated as a result of decontamination and decommissioning of nuclear facilities.

LOW-LEVEL WASTE MANAGEMENT

Regulatory Development
NRC continued its low-level waste regulatory
development effort in 1979 with the objective of
publishing a draft regulation (10 CFR 61) on low-level
waste disposal. An environmental impact statement(EIS) will be prepared to support the rulemaking action. Work was also continued on supporting
regulatory guides and staff positions.
The draft regulation as currently envisioned willconsist of basic performance objectives applicable tothe disposal of low-level waste on land by various
methods. These objectives will be met by establishingappropriate requirements for siting a disposal facility
and assuring adequate operations site closure anddecommissioning and adequate institutional arrangements. Technical details specific to the individual disposal techniques of shallow-land burial
and other alternative disposal methods will be contained in appendices to the regulation and in
regulatory guides. A preliminary draft of the regulation was completed in 1979 and made available to awide cross section of persons for informal review. The
draft will be made available to the public in 1980.
The regulatory guides associated with the regulation
are also under development and are currently envisioned to cover waste form and content; site designand operations; site monitoring and surveillance; site
closure, stabilization, and post-operational care;standard contents for license application and environmental report; records and reports; and funding.
In addition to the above work, NRC has contracts
with various organizations to develop a base of supporting technical information. Contractual studies areunderway in such areas as systems analysis, waste
classification, and volume reduction. The systems
analysis contractor is developing models for analyzing
radioactive waste disposal by shallow-land burial. The

This "fish pole" radiation survey meter penn its inspectors to accuratelyassess the radioactivity of low-level waste material in trenches prior to burial. Containers have just been delivered anddumped by trucks in background and will be covered by earthmoving equipment as soon as radiation levels and distribution havebeen recorooo.

NRC's work in regulatory development in 1979 has
been focused on development of requirements that canapply to a broad range of disposal alternatives. It has
become increasingly dear to the NRC during 1979 thatalternative disposal methods are critically needed anda regulatory base should be put in place in timely manner.

Licensing
NRC continued its licensing activity in low-level
waste management in 1979. The NRC license for
disposal of special nuclear material (SNM) at Hanford,
Wash., was renewed in November 1979. This license
was closely coordinated with the State of Washington
and contains many upgraded requirements for operations at the site.
An environmental assessment was continued in 1979
for decommissioning of the Sheffield, Ill., facility andshould be completed in 1980. The licensee applied foran expansion and continued operation of the site.

150

Low-level waste containers that contain hi~er-activity materials(or that emit higher levels of radioactivity) than those dumped instandard low-level disposal trenches are deposited in trenches orcontainment holes which offer greater depths and heavier

shielding. Two types of such containments are shown here: (left) a
reinforced concrete lined pit, and (right) a narrow, deep trenchshielded by the filled barrels along the top. Both such containments are in protected, posted areas at a supervised site.

However, the licensee subsequently petitioned towithdraw the renewal and expansion application. The
Atomic Safety and Licensing Board approved thewithdrawal of the expansion request but the renewal will be subject to hearings. The applicant's
withdrawal of the operating/expansion application
was based on recognized technical problems for which
the solution proposed by the licensee was not acceptable to the NRC. In addition to the above, five licenseamendments were granted for existing sites.
Since two of the previous six commercial disposal
operations have closed, (West Valley, N. Y. and Maxey
Flats, Ky.) and the Sheffield, Ill., disposal operation iseffectively closed, only three commercial operations
currently exist (Barnwell, S.C.; Beatty, Nev.; andRichland, Wash.). Thus the present disposal capacity
is primarily located in the West and Southeast andrepresents an undesirable regional imbalance. The
waste from reactors and other waste generators
located in the Northeast and Midwestern United States
must be transported either to the Southeastern UnitedStates or to the West. .
A number of significant events occurred in 1979 that
affect low-level waste disposal operation. It became
obvious that more attention should be paid to decontamination and decommissioning wastes, from theviewpoint of low-level waste disposal operations.

Some of these activities pose unique problems, such as
the TMI waste and the waste from the decontamina.
tion of the Dresden I reactor. It also became obviousthat further work is required for liquid scintillation
waste. The State of South Carolina decided in 1979
not to accept any more shipments of this type of waste,
and the waste must presently be shipped to thedisposal operations in the West. NRC is investigating
various alternatives for the treatment and disposal ofthis type of waste.
Lastly, it became obvious that NRC must take amore active role in upgrading packaging requirements
and waste form for certain types of waste and increaseinspection and enforcement of existing regulations
covering the shipment of waste, For example, a fireoccurred on a truck containing waste packages at theBeatty, Nev., site and large volumes of free-standing
liquids were found upon inspection of packages ofsolidified wastes received at various low· level waste
disposal sites. As a result of such events, the governorsof the three States having commercial low-level waste
disposal operations sent a joint communique dated July 10, 1979, to NRC demanding action by NRC andthe Department of Transportation to improve packaging requirements and increase inspection and enforcement of existing regulations. In response, NRC issued abulletin to all licensees stressing the need to give===================================================151
careful attention to the packaging and transportationof waste and instituted action, with the cooperation of
the States and the DOT, to inspect shipments on amore frequent basis and take more stringent enforcement actions. (See also Chapter 4.)
As mentioned above, a severe regional imbalance
has emerged from the locations of today's low-level
waste burial grounds. This imbalance was aggravatedin 1979 when two of the sites closed and then reopened
and a curtailment was placed on the amount of wastethat could be received at the third site. As a result,
NRC went on record to state its judgment that lowlevel waste disposal is the responsibility of the States,
for the States receive the benefits of the operations
which generate the waste. NRC has worked with anumber of States in 1979 and will continue to do so in
1980, to help the States explore the possibility of
establishing new sites. The NRC effort took the formof assistance in setting forth licensing and regulatory
requirements; however, NRC cannot promote theopening of new sites. This is a responsibility of theStates, with assistance available from the Departmentof Energy should the States request such assistance.

Technical Assistance toAgreement States
NRC has provided technical assistance to Agreement States in the licensing and regulation of low-level
waste disposal operations in their jurisdiction. NRC
has provided in 1979, assistance to the State ofWashington as part of their renewal action for theState disposal license at Richland. In addition, NRC
has provided,and will provide in 1980, assistance to
the State of Kansas in evaluating an application for a
new disposal site license at Lyons. Technical assistance
was also given to the State of Nevada in 1979, and
NRC is expecting to provide further assistance to
Nevada in 1980 regarding renewal of the State license
for the Beatty site. The NRC technical assistance supplements the State's resources and assures that thetechnical criteria used to license and regulate a lowlevel waste disposal operation in an Agreement Stateare compatible with the criteria used to license andregulate a low-level waste disposal operation under
NRC's jurisdiction. In 1979, NRC worked with theStates of South Carolina, Nevada and Washington todevelop and implement new requirements at existingsites to upgrade and define acceptable waste forms.

URANIUM RECOVERY MANAGEMENT
Regulatory Development
NRC continued its uranium recovery regulatory
development effort in 1979 with the objective of
upgrading its regulations for uranium milling in 1980.
The NRC published a draft generic environmental im-

pact statement (GElS) in April 1979 covering the U.S.
uranium milling industry to the year 2000, with particular emphasis on mill tailings. In addition, NRC
published draft regulations in August 1979, derivingfrom the environmental statement, and conducted extensive public meetings on the proposed regulations.
The final GElS and the final regulations are expected
to be published in 1980.
The proposed regulations cover radioactive airborne
emissions during operation, mill tailings disposal,
decommissioning of mill structures and sites, supplementary institutional and procedure requirements,
implementation of proposed requirements at existingsites, and heap leaching and small processing sites.

Licensing
NRC continued its licensing effort in 1979. Twelve
new uranium recovery facilities were licensed and one
facility license was renewed. In addition, five majoramendments were issued based upon licensee requests
for facility modifications. There were 15 uranium
mills, 5 heap leach/ore buying stations, 2 solution mining operations and 16 research and development(R&D) operations under NRC license in 1979.
Similar facilities exist in Agreement States. All thesetypes of facilities are expected to grow numerically inthe future. It is currently projected that in 1981 there
will be 22 operating mills, 8 heap leach operations andore buying stations, 6 commercial scale solution mining operations, and 23 R&D operations under NRC
jurisdiction. A similar growth is expected in thenumber of these types of operations in Agreement
States. Thus, the NRC and Agreement State workloadin this area will experience a substantial growth in thenext few years.

Technical Assistance to Agreement States
During 1979, NRC provded technical assistance to
the States of Washington, Oregon, Colorado, New
Mexico, Arizona, California and Nevada in the licensing and regulation of uranium recovery operations
under Agreement State jurisdiction. A total of six project reviews were completed. These reviews covered
uranium mills, heap leach operations, solution mining
operations, and R&D operations. The NRC assistanceassures that the technical criteria used to license andregulate uranium recovery operations in Agreement
States are compatible with those criteria used tolicense and regulate similar operations under NRC
jurisdiction.
The Uranium Mill Tailings Radiation Control Act of1978 was amended in 1979 to provide further clarification of the NRC/ Agreement States interface withrespect to the licensing and regulation of mill tailings.
The Commission will continue to license tailings in
non-Agreement States and the Agreement States willcontinue to license the mill tailings under State

152=================================================
Technical assistance to the Agreement States by
NRC will continue to cover non-routine safety and environmental assessment. For example, a tailings impoundment failure occurred at the United Nuclear
Corporation uranium milling operation at Church
Rock, N.M., on July 16, 1979. New Mexico is an
Agreement State and the milling complex was licensed
by the State in May 1977. Estimates of the amount oftailings released have varied, but it appears that about
100 million gallons of acidic tailings solutions and1,100 tons of tailings solids escaped from the tailings
impoundment area before the break in the dam could
be closed. The State of New Mexico requested
technical assistance from NRC and NRC personnel
were dispatched to the site to aid the State. Extensive
technical studies and analyses were also performed by
NRC. Technical assistance to the State of New Mexico
will continue to be provided by NRC in 1980.

NRC Assessment of
DOE Remedial Action Plans
A State inspector and a State Agreements program reviewer ex~
amine a waste burial trench at Barnwell, S.C. Low-level radioactive wastes are deposited in such trenches and covered withbackflll. Only three low-level waste burial sites are now operatingin the United States. Barnwell is the only site in the eastern part ofthe country. The other two sites are at Hanford, Washington andBeatty, Nevada.

jurisdiction. NRC will provide technical assistance to
the States in carrying out their responsibilities underthe Act.

NRC initiated in 1979 its evaluation of DOE
remedial action plans for inactive sites. This will be afive year program which implements NRC's part of
Title I of the Uranium Mill Tailings Radiation Control
Act of 1978. DOE is responsible for remedial action at
21 inactive mill tailings sites and one other former ore
processing site as specified in the Act. NRC is required
to review DOE's proposed remedial actions anddetermine whether the remedial action plans are
acceptable.

7
Inspection and
Enforcement

New emphasis was given in 1979 to direct
NRC inspection of design, analytical andother technical activities of contractors.

During 1979, the NRC continued to implement theplan calling for resident inspectors at each operating
power reactor plant, at those plants in the later stagesof construction, and at selected fuel cycle facilities.
The accident at Three Mile Island (TMI) led to a decision to increase the number of resident inspectors to alevel of one inspector for each unit at a multireactor
site. Single unit sites will have two resident inspectors.
Ry December 31, 1979, 60 inspectors were stationed as
residents at 48 power reactor and fuel facility sites.
Table 1 provides a listing of these sites. This additional
effort has required an increase in the number of personnel from a staff ceiling of 715 in 1979 to 861 in 1980
for the NRC Office of Inspection and Enforcement. By
the end of fiscal year 1980 there will be 157 resident inspectors on site compared to the original goal of 76.
The reactor training provided for operations inspectors will be increased from a minimum of seven weeks
to 10 weeks during 1980, with additional simulator
and special plant observation training.
TMI impacted heavily on the planned inspection
program. Special teams were sent to all operatingpressurized water nuclear plants to review withlicensee management the actions required as a result of
the TMI accident. Review groups were formed to
study the TMI accident and the lessons learned from itthat would affect future inspection programs. An
augmented 24-hour surveillance program wasestablished at TMI that has required staffing support
from all five NRC regional offices. From April throughJuly of 1979, a 24-hour watch was established in eachregion and at the NRC Operations Center in Bethesda,
Maryland, to provide the capability for responding
immediately to any incidents or accidents. A direct"hotline" telephone system was installed in the Operations Center. This provides a direct line to each
operating reactor power plant and all fuel processing
facilities in the country. The system provides conference call capability between the NRC Operations

Center, a plant, and the regional NRC office. In
August, the 24-hour duty Officers in the regions were
replaced by a communications system connected
directly to the NRC Operations Center where 24-hour
duty officer coverage is maintained. All calls toregional offices during non-duty hours are now
diverted to the NRC Duty Officer at the Operations
Center, who can promptly respond to the situation.
As a result of these actions associated with the TMI
accident and related inspections, the number ofroutine inspections in 1979 was less than originallyplanned. Table 2 summarizes the inspections conducted during fiscal year 1979.
One or more noncompliance items were found in 33
percent of the more than 6,000 inspections and in 36
percent of the 121 investigations. The more severe
sanctions imposed on licensees for failure to comply
with NRC requirements included nine civil penalties
and three orders to "cease and desist" operations, orfor modifications, or suspensions of licenses (see Tables
4 and 5).

THE INSPECTION PROGRAM
The inspection and enforcement program is directedby NRC's Office of Inspection and Enforcement (IE),with a headquarters staff located in Bethesda,
Maryland, and a field staff deployed in NRC's five
regional offices located in or near Philadelphia, Atlanta, Chicago, Dallas, and San Francisco. About 80 percent of the total office on-board staff of 730 is assigned
to the regions.
The objectives of inspections are:
• To determine whether licensees are complying
with NRC requirements.
• To identify conditions that may adversely affectpublic health and safety, the common defense

154

Table 1. Sites Manned by Resident Inspectors During 1978 and 1979
Facility'" Arkansas Nuclear Plant
Beaver Valley Power Station
Bellefonte Nuclear Plant"'Browns Ferry Nuclear Plant
Brunswick Steam Electric Plant
Callaway Plant
Calvert Cliffs'" Comanche Peak Steam Electric
Station
Davis-Besse Nuclear Power Station'" Donald C. Cook Plant"'Diablo Canyon Nuclear Power Plant'" Dresden Nuclear Power Station'" Edwin 1. Hatch Plant
Fort St. Vrain Nuclear Station
Hartsville Nuclear Power Plant"'Indian Point Station
Joseph M. Farley Nuclear Plant
LaSalle County Nuclear Station
Limerick Generating Station
Marble Hill Plant"'Midland Nuclear Power Plant"'Millstone Nuclear Power Station
North Anna Power Station"'Oconee Nuclear Station
Palisades Nuclear Power Station
Palo Verde Nuclear Station"'Peach Bottom Atomic Power
Station"'Prairie Island Nuclear Plant
Quad Cities Station
Rancho Seco Nuclear Station'" Salem Nuclear Generating Station"'San Onofre Nuclear Station
Seabrook Nuclear Station
Sequoyah Nuclear Power Plant
Shoreham Nuclear Power Station
South Texas Nuclear Project
Summer Nuclear Station'" Surry Power Station'" Susquehanna Steam Electric Station"'Trojan Nuclear Plant
Turkey Point Station
Washington Nuclear #2"'Watts Bar Nuclear Plant
William H. Zimmer
Nuclear Power Station
Zion Nuclear Plant"'B&W-Apollo & Leechburg'" '"(Fuel Facility)"'Westinghouse-Cheswick'" '" (FuelFacility)"'Nuclear Fuel Services (Fuel Facility)

Location
Russelville, Ark.
Shippingport, Pa.
Scottsboro, Ala.
Decatur, Ala.
Southport, N.C.
Fulton, Mo.
Lusby, Md.
Glen Rose, Tex.
Oak Harbor, Ohio
Bridgman, Mich.
San Luis Obispo, Cal.
Morris, Ill.
Baxley, Ga.
Platteville, Colo.
Hartsville, Tenn.
Indian Point, N.Y.
Dothan, Ala.
Seneca, Ill.
Pottstown, Pa.
Madison, Ind.
Midland, Mich.
Waterford, Conn.
Mineral, Va.
Seneca, S.C.
South Haven, Mich.
Winter burg, Ariz.
Peach Bottom, Pa.
Red Wing, Minn.
Cordova, Ill.
Sacramento, Cal.
Salem, N.J.
San Clemente, Cal.

LicenseeArkansas Power & Light Co.
Duquesne Light Co.
Tennessee Valley Authority
Tennessee Valley Authority
Carolina Power & Light Co.
Union Electric Co.
Baltimore Gas & Electric Co.
Texas Power & Light, Dallas Power& Light, Texas Electric Service
Toledo Edison Co.
Indiana & Michigan Electric Co.
Pacific Gas & Electric Co.
Commonwealth Edison Co.
Georgia Power Co.
Public Service Co. of Colorado
Tennessee Valley Authority
Consolidated Edison Co.
Alabama Power Co.
Commonwealth Edison Co.
Philadelphia Electric Co.
Public Service of Indiana
Consumers Power Co.
Northeast Nuclear Energy Co.
Virginia Electric & Power Co.
Duke Power Co.
Consumers Power Co.
Arizona Public Service Co.
Philadelphia Electric Co.

Seabrook, N.H.
Daisy, Tenn.
Suffolk County, N. Y.
Bay City, Tex.
Broad River, S.C.
Gravel Neck, Va.
Berwick, Pa.
Prescott, Ore.
Florida City, Fla.
Richland, Wash.
Spring City, Tenn.
Moscow, Ohio

Northern States Power Co.
Commonwealth Edison Co.
Sacramento Municipal Utility District
Public Service Electric & Gas Co.
Southern California Edison Co. & San
Diego Gas & Electric Co.
Public Service Co. of N.H.
Tennessee Valley Authority
Long Island Lighting Co.
Houston Lighting & Power Co.
South Carolina Electric & Gas Co.
Virginia Electric & Power Co.
Pennsylvania Power & Light Co.
Portland General Electric Co.
Florida Power & Light Co.
Washington Public Power Supply System
Tennessee Valley Authority
Cincinnati Gas & Electric Co.

Zion, Ill.
Apollo, Pa.

Commonwealth Edison Co.
Babcock & Wilcox Co.

Parks Township, Pa.

Westinghouse Electric Corp.

Erwin, Tenn.

Nuclear Fuel Services, Inc.

'" Assigned during calendar year 1978.
'" "'Inspector stationed at Apollo, Pennsylvania, acts as inspector on a rotating basis at B&W's Apollo and Leechburg facilities andWestinghouse's Cheswick facility.

155

and security, the environment or the safeguarding of nuclear materials and facilities.
• To provide information to assist in developing abasis for issuance, denial, or amendment of an
authorization, permit or license.
• To determine whether licensees and their con~tractors and suppliers have implemented adequate quality assurance programs.
When an inspection or investigation discloses events orconditions that present a potential or actual threat to
public health and safety, the environment, or thesafeguarding of nuclear materials and facilities, the
NRC takes prompt action and routinely communicateswith other parts of government, licensees and the
public.
During fiscal year 1979, 174 new inspection procedures and! or instructions were issued and 123 wererevised. In the area of construction inspection, for example, 22 extensively revised inspection procedures
pertaining to welding were issued.

Reporting Defects and Noncompliance
On June 6, 1977, the NRC published in the Federal
Register a regulation (10 CFR Part 21) setting forth therequirements for implementing Section 206 of theEnergy Reorganization Act of 1974. Individual directors or responsible officers of a firm involved in the
nuclear industry are required to report noncompliancewith NRC regulations or the existence of defects which
could create a substantial health and safety hazard.
Any such person who knowingly and consciously fails
to provide the required reports to the NRC is subject toa civil penalty not to exceed $5,000 for each failure
and a total amount not to exceed $25,000 within any
30-day period. The regulation became fully effective
on January 6, 1978.

About 150 Part 21 Reports have been received by
the NRC since the regulation became effective. The
reports are reviewed to assess the reported deficiency,
the adequacy of the proposed corrective action and thepossibility of generic problems. IE inspectors ensurethat appropriate followu'p actions are taken.

Types of Inspections
NRC's inspections are of two basic types: routine
and reactive. In routine inspections, NRC inspectors
concentrate on determining the effectiveness of quality
assurance systems by direct observation and verification of licensee activities, and by reviewing procedures, checking records, interviewing people, and,
where appropriate, making direct measurements .
. Reactive inspections are conducted in response to information received by NRC regarding conditions or
events affecting licensed facilities or material under
NRC jurisdiction. Such information may come fromroutine NRC inspections; from an applicant, licensee,
contractor or supplier; or from licensee employees orother members of the public.
Inspections cover the entire range of NRC licensed
activities. Reactor-related inspections cover all phasesof nuclear power plants (preconstruction activities,
construction, preoperational testing and startup,
operation, shutdown and decommissioning) andsimilar phases of research and test reactors. In addition, NRC inspects the quality assurance programs ofcontractors and vendors who supply safety-related
equipment, components and services to power reactors
under construction or in operation.

Licensee, Contractor and Vendor Inspection
ProgramApproximately one-half the work associated withconstructing a nuclear facility is accomplished off~site.

Table 2. Inspections Conducted in Fiscal Year 1979
Num~er of

ProgramPower Reactor Construction
Operating Power Reactors
Other Reactors
Fuel Facilities
Materials
Vendors
Safeguards

Licenses

114
70
94
39
8,586
248
203

Number ofInspections

1,787
1,761
93
203
1,976
228
526

156

Starting above and reading clockwise, an inspector is shown during
an inspection of pump vendor facilities, as he reads and records
data during reactor coolant pump test, checks gas-metal arc
welding being performed on a pump rotor, observes machining ofpump rotor housing, watches vendor technician balance a pumprotor, and (below) reviews records of material identification andtraceability .

157This includes facility design and the fabrication ofcomponents of safety-related systems. Inspections of
nuclear steam system suppliers, architect-engineersand vendors of safety-related components are performed by NRC's Licensee, Contractor and Vendor
Inspection Branch (LCVIB) inspectors, located in the
Region IV (Dallas) office. During fiscal year 1979,
some 250 inspections were performed by the 21
LCVIB inspectors. Approximately 30 percent of these
inspections were special reactive inspections involving
component fabrication or design-related problems.
During the coming year, a modest shift in inspection
emphasis is expected in the LCVIB. Activities experiencing change will include:• Performing more reactive inspections.
• Redirecting emphasis toward the inspection oftechnical activities performed by contractors.
• Followup on Part 21 Reports, Bulletins and Circular issues.
• Inspecting and witnessing environmental qualification of electrical, instrumentation and control
equipment.
• Inspecting design and analytical work performedby licensee contractors.

Performance Appraisal Program
During fiscal year 1979, five licensee management
appraisal inspections and one IE program appraisal
inspection series (pertaining to surveillance testing)were completed. Nine management appraisal inspections and four IE program appraisal inspections areplanned for fiscal year 1980. Objectives of the program are to:• Evaluate performance of utility management.
• Analyze effectiveness of the NRC inspection program.
• Confirm objectivity of NRC inspectors.
Three Performance Appraisal Team (PAT) inspectors participated in the IE investigation of the TMI accident; PAT inspectors also participated in other investigations and special inspections.

Independent Measurement/Verification
Program
IE has increased its efforts associated with directverification of licenseel contractor activities during theconstruction phase. NRC periodically uses contractors
to perform non-destructive testing activities, and, inAugust 1979, selected a contractor to perform destructive testing of selected materials used in safety-related
structures and systems. Continued effort in these areasis planned for fiscal year 1980.

Inspections related to nuclear materials include inspection of the construction and operation of uranium
mills; fuel fabrication, processing and reprocessing
plants; waste disposal facilities; and the industrial,
educational and medical uses of radioactive material.
NRC· inspections also include measures for safeguarding nuclear material from theft and sabotage, forphysical protection of reactors and fuel cycle facilities,
and for transportation of nuclear materials.
The number of inspections carried out during fiscal
year 1979 (ending September 30) for each of these activities is shown in Table 2.

Government-Industry Efforts
The NRC inspection program is based on thepremise that the licensee is responsible for carrying out
licensed activities safely and in compliance with NRC
requirements. NRC determines whether the licensee
has established the management control systems
necessary to meet regulatory responsibiliti~s. The inspection pattern for large, complex nuclear facilities ispyramidal, with each level of activity verified, inspected or audited by those above. The NRC inspection effort is essentially the apex of the pyramid, i.e.,
NRC performs the last in the series of inspections andaudits conducted by many different groups. Since
NRC inspection manpower is usually far less than thatof licensees and contractors, NRC inspectors cannot
inspect all components and activities; thus, they probe
the "pyramid" to determine whether the licensees' andcontractors' activities are properly performed. In addition, the IE inspection program provides for independent effort by NRC inspectors whenever the inspector determines such action is necessary.

Inspection Activities Resulting from TMI
Shortly after the Three Mile Island (TMI) accident,
a series of Inspection and Enforcement (IE) Bulletinswere issued to all operating power reactor licensees addressing the early lessons learned. The IE Bulletinsprovided licensees with information about the series of
events that occurred at TMI and directed each licensee
to make changes to certain equipment and operatingprocedures consistent with the reactor design, and to
conduct special operator training. In response to the
Bulletins, licensees provided details for completion of
immediate actions and plans for completion of longer
term actions. Special follow-up inspections were conducted to verify that licensees had taken appropriate
action.
During the period April 18-23, 1979, six speciallytrained NRC teams visited all operating pressurizedwater nuclear power plants, -except those designed byBabcock and Wilcox, designer of the TMI plant. These

158

Pressurized water reactor control room simulators, such as this one
at the Tennessee Valley Authority's Sequoyah Nuclear Power
Plant, came into greaUy increased we m 1979. The training of
new inspectors to accommodate the growing NRC resident inspector

program brought increased student loads to such facilities eady inthe fiscal year, and the training was intensified even more in thelater months as deficiencies in operator training highlighted by theThree Mile Island accident became apparent.

te"ains reviewed and discussed with licensee operations
personnel and station management the TMI accident
chronology and licensee actions that had been specified in the IE Bulletins. For the Babcock and Wilcox
designed facilities, the resident inspector, withassistance from regional-based inspectors, conducted
this special briefing.

was significant. Efforts expended by inspectors at the
TMI site, the expedited assignment of inspectors to all
Babcock and Wilcox power reactor facilities and therequirements imposed by the need for special inspections of all operating power reactor facilities caused athinner coverage, and in many cases deferral or deletion of portions of the routine inspection program activities.

By April 2, following the TMI accident, resident inspectors had been assigned to all operating Babcock
and Wilcox designed plants where resident inspectors
had not previously been assigned. In addition to
responsibilities normally assigned resident inspectors,
inspectors at these sites performed additional inspections to assure plant safety in light of the events at
TMI.
.

Impact of TMI on Inspection Program
The impact on the routine inspection program forthe first several months following the TMI accident

On a continuing basis, increased emphasis has beenplaced on identifying isolated plant problems andgeneric issues and managing theh; resolution. To accomplish this, the headquarters staff has been
augmented with a group of highly specialized systems
engineers whose responsibilities include more in-depth
review and follow-up on plant events.
Long-term inspection program changes to reflectlessons learned from the TMI accident are still invarious formative stages. Specific problems 'requiring
program changes have generally been diagnosed. Program modification, implementation and attendant

159

process evaluations have been done to the extent possible for changes that represent an expansion of current
programs, such as resident inspection.
Studies to evaluate certain major changes in emphasis of the inspection program have been initiated todetermine the effectiveness and efficiency of thesechanges as implemented. The results of the IE Special
Review Group on lessons learned from Three Mile
Island have provided a basis for the integration oflessons learned into the current inspection program.

Resident Inspector Program
During 1979, the NRC made further progress in the
program to station inspectors full time at the sites ofnuclear power reactors and major fuel cycle facilities.
Experience with resident inspection results andlicensee events and actions have led to plans for further expansion of the resident inspector program. The
program is being accelerated in consonance with the
President's message of December 7, 1979 on the
Kemeny Commission Report. Steps to upgrade its effectiveness also are being taken in response to recommendations in a General Accounting Office reportissued to Congress in November 1979. As noted above,
approval has been given to assigning, in addition to
the site resident inspector, resident inspectors tonuclear power reactor plant units (many sites havemore than one unit). The total number of resident in·spectors at any site will generally equal the number ofunits at that site, with a minimum of two inspectors
per site. This augmented coverage will provide additional safety assurances through increasing NRC
presence, including the number of independent
observations of licensee safety-related activities andequipment.
By December 31, 1979,60 inspectors were deployed
to the sites of 45 nuclear power stations-including
several' power reactor plants under construction-andof three fuel facilities. By June 1980, each site with anoperating or preoperational reactor should have atleast one resident inspector. Each such site is expected
to have its full complement of at least two inspectors
by September 30, 1980, at which time some 130 resident inspectors will be deployed at 60 sites.
Thereafter, resident inspectors will be assigned toreactors as they reach the pre-operational stage.
The NRC also is assigning resident inspectors to sites
where nuclear plant construction is in the final stage.
Further, resident inspectors will be assigned to sites
where problems are evident in earlier stages of plant
construction.

number and significance. The NRC's Office of Inspection and Enforcement has issued Bulletins since 1971,
Circulars since 1976 and Information Notices for the
first time in 1979.
The IE Bulletin is used to notify licensees of specifica~tions to be taken. It usually requires that thelicensees provide a report to the NRC describing theactions they take in response to the Bulletin. The
Bulletin addresses matters of concern or events related
to reactor safety, material safeguards, radiologicalsafety or environmental protection.
Bulletins usually, although not always, require theaction on a one-time only basis. However, Bulletinsare not intended to substitute for new or revised
license conditions or requirements. If a licensee refuses
to perform an action set forth in the Bulletin, the requirement for the action may be imposed on the licensee
by an Order.
Particular considerations which might require theissuance of a Bulletin include events in which the safety significance is of such a magnitude as to result in an
immediate impact on all of a certain type of licensee.
The Three Mile Island accident represents such an
event, and it was addressed by multiple Bulletins.
Other considerations include events having a potential
generic problem impact and where the event requires
action by a particular class of license or permit holder.
The IE Circular is used to notify licensees of actions
which the NRC recommends be taken. These matters
are generally of lesser significance than those address-

Bulletins, Circulars and Information Notices
During 1979, the NRC's issuance of Bulletins, Circulars and Information Notices was increased both in

NRC resident inspe<?tor checks a weld in a reactor vessel thermal
sleeve at the Susquehanna Steam Electric Station at Berwick, PA.

160

Table 3. IE Bulletins, Circulars, and Information Notices Issued in 1979
BULLETINS

Bulletin No.

Subject

Date Issued

Issued to79-01

Environmental
Qualification ofClass IE
Equipment

2/8/79All power reactor
facilities withan OL or CP

79-01A

Environmental
Qualification ofClass IE Equipment(Deficiencies in theEnvironmental
Qualification of
ASCO Solenoid Valves

6/6/79All power reactor
facilities withan OL or CP

79-02

Pipe Support Base
Plate Designs
Using Concrete
Expansion Anchor
Bolts

3/2/70All power reactor
facilities withan OL or CP

79-02(Rev. 1)

Pipe Support Base
Plate Designs
Using Concrete
Expansion Anchor Bolts

6/21179

All power reactor
facilities withan OL or a CP

79-02
(Rev. 1)

Pipe Support Base Plate
Designs Using Concrete(Supplement 1)

8/20/79

All power reactor
facilities with an
OL or a CP

79-03

LongitudinalWelds Defects
in ASME SA-312
Type 304 StainlessSteel Pipe Spools
Manufactured byYoungstown Welding
and Engineering Co.

3/12/79

All power reactor
facilities with an
OL or CP

79-04

Incorrect Weights
for Swing Check
Valves Manufactured
by Velan Engineering
Corporation

3/30/79

All power reactor
facilities with an
OL or CP

79-05

Nuclear Incident
at Three Mile Island

4/2/79All power reactor
facilities withan OL and CP

79-05A

Nuclear Incident
at Three Mile Island

4/5/79All B&W powerreactor facilities
with an OL

79-05B

Nuclear Incident atThree Mile Island

4/21179

All B&W powerreactor facilities
with an OL

79·05C&06C

Nuclear Incident atThree Mile Island Supplement

7/26/79

To all PWR powerreactor facilities
with an OL

161

BULLETINS

Date Issued

Issued to

Review ofOperational Errors
and System Misalignments
Identified During
the Three Mile
Island Incident

4/11/79

All pressurizedwater powerreactors with an
OL except B&W
facilities

79-0BA

Review ofOperational
Errors and System
Misalignments
Identified During
the Three Mile
Island Incident

4/14/79

All pressurizedwater powerreactor facilities
of Westinghouse
design with an OL

79-0BA(Rev. 1)

Review 'of Operation.!!.lErrors and System
Misalignments
Identified During
the Three Mile
Island Incident

4/18/79

All pressurizedwater powerreactor facilities
of Westinghouse
design with an OL

79-0BB

Review ofOperational
Errors and System
Misalignments
Identified During
the Three Mile
Island Incident

4/14/79

All Combustion
Engineering designed
pressurized Water,
power reactor
facilities with
an OL

79-07

Seismic Stress
Analysis of
Safety-Related
Piping

4/14/79

All power reactor
Facilities with an
OL or CP

79-08

Events Relevant
to BWR Reactors
Identified
During Three Mile
Island Incident

4/14/79

All BWR powerreactor facilities
with an OL

79-09

Failures of GE
Type AK-2 Circuit
Breaker in Safety
Related Systems

4/17/79

All power reactor
facilities with an
OL or CP

79-10

Requalification
Training Program
Statistics

5/11/79

All power reactor
facilities with an
OL

79-11

Faulty Overcurrent
Trip Device inCircuit Breakers
for Engineered
Safety Systems

5/22/79

All power reactor
facilities with an
OL or a CP

79-12

Short Period
Scrams at BWR
Facilities

5/31/79

All GE BWR
facilities withan OL

78-12BA Typical Weld
Material in Reactor
Pressure Vessel
Welds

3/19/79

All power reactor
facilities withan OL or CP

Bulletin No.

Subiect

79-0B

162

Table 3. IE Bulletins, Circulars, and Information Notices Issued in 1979-Continued
BULLETINS-Continued

Bulletin No.

Sub;ect

79-13

Cracking inFeedwater System
Piping6/25179

All PWRs with an
OL for action;
all BWRs with a
CP for information

79-14

Seismic Analyses
for As-Built
Safety-Related
Piping System

6/2179All power reactor
facilities with an
OL or a CP

79-15

Deep Draft Pump
Deficiencies

7/11179

All power reactor
licensees with a
CP and/or OL

79-16

Vital Area AccessControl

7/26/79

All holders of andapplicants for 0 L

79-17

Pipe Cracks in Stagnant
Borated Water Systems at
PWR Plants

7/26179

All PWR's OL

79-18

Audibility Problems
Encountered on Evaluation

8/7179All power reactor
facilities with an OL

79-19

Packaging Low-Level
Radioactive Waste forTransport and Burial

8/10/79

All power and research
reactors with OLs, fuelfacilities except uranium mills, andcertain materials licensees.

79-20

Packaging Low-Level
Radioactive Waste forTransport and Burial

8/10/79

All materials licensees
who did not receive
Bulletin No. 79-19

79-21

Temperature Effects
on Level Measurements

8/13/79

All PWRs with an OL

79-22

Possible Leakage of Tubes
of Tritium Gas in Timepieces
for Luminosity

9/5179To each licensee who
receives tubes of tritium
gas used in timepieces
for luminosity

79-23

Potential Failure of
Emergency Diesel Generator
Field Exciter Transformer

9/12179

All power reactor
facilities with an
OL or a CP

79-24

Frozen Lines

9/27179

All power reactor
facilities which have
either OLs or CPs andare in the late stageof construction

Date Issued

Issued to

CIRCULARS

Circular No.

Sub;ect

79-01

Administration of
Unauthorized
Byproduct Material
To Humans

Date Issued
11 12179

Issued to
All holders oflicenses except
teletherapy medical
Licenses and EachRadiopharmaceu tical
Supplier

163

Circular No.

Subject

Date Issued

Issued to

79-02

Failure of 120
Volt Vital AC
Power Supplies

2/16179

All holders ofreactor OLs and
CPs

79-03

Inadequate Guard
Training
Qualification andFalsified Training
Records

2/23179

All holders ofand applicants for
special nuclear
material licenses
in safeguards
Group

79-04

Loose Locking Nut
on Limitorque Valve
Operators

3/16179

All holders ofreactor 0 Ls or
CPs

79-05

Moisture Leakage
in Stranded Wire
Conductors

3/20179

All holders ofreactor OLs or
CPs

79-06

Failure to Use
Syringe andBottle Shields
In Nuclear Medicine

4/19179

All holders of
medical licenses
except teletherapy
licensees Issued to

79-07

Unexpected Speed
Increase of
Reactor Recirculation
MG Set Resulted inReactor Power
Increase

5/2179All holders of
BWR OLs or CPs

79-08

Attempted
Extortion Low
Enriched Uranium

5/18179

All fuel
facilities
licensed by NRC

79-09

Occurrences of
Split or Punctured
Regulatory Diaphrams
in Certain Self
Contained Breathing
Apparatus

6/22179

All materials
priority I, fuelCycle andOperatingreactor licenses

79-10

Pipefittings
Manufactured from
Unacceptable
Material

6/26179

All power reactor
licensees with a
CP and/or OL

79-11

Designl Construction
Interface Problem

6/27179

All applicantsfor, and holders
of Power Reactors
CPs

79-12

Potential Diesel
Generator
Turbocharger
Problem

6/28179

All power reactor
operation
facilities and all
utilities having a CP

79-13

Replacement ofDiesel Fire Pump
Starting
Contactors

7/10179

All power reactor
Operations
facilities andall utilities
having a CP

79-14

Unauthorized Procurement andDistribution of XE-133

7/13179

All medical licensees
except teletheraphy
medical licensees andto all radiopharmaceu
tical suppliers164

Table 3. IE Bulletins, Circulars, and Information Notices Issued in 1979-Continued
CIRCULARS
Circular No.

Subject

Date Issued

Issued to

79-15

Bursting of High Pressure Hoseand Malfunction of Relief
Valve "0" Ring in Certain SelfContained Breathing Apparatus

8/8/79All materials Priority I,
fuel cude and operating
power reactor licensees

79-16

Excessive Radiation Exposures
To Members of the general Public
and a Radiographer

8/16/79

All radiographylicensees

79-17

Contact Problem in SB-12
Switches on General ElectricCompany Metal Clad Circuit
Breakers

8/14/79

All power reactor
licensees with a CP and/
or OL

79-18

Proper Installation of
Target Rock Safety-Relief

9/10179

All holders of powerreactors OLs and CPs

79-19

Loose Locking Devices
on Ingersoll-Rand Pumps

9/13/79

All holders of powerreactors OLs and CPs

79-20

Failure of GTE Sylvania
Relay, Type PM Bulletin
7305, Catalog 5&12-11-AC

9/24/79

All holders of powerreactors OLs and CPs

INFORMATION NOTICES
Information
Notice No.
79-01

79-02

79-03

79-04

79-05

79-06

79-07

79-08

Subject
Bergen-Paterson
Hydraulic Shockand Sway Arrestor
Attempted
Extortion of Low
Enriched Uranium
Limitorque Valve
Geared Limit
Switch Lubricant
Degradation ofEngineered
Safety Features
Use of Improper
Materials in
Safety-Related
Components
Stress Analysis
of Safety-Related
PipingRupture of
Radwaste Tanks
Interconnection ofContaminated
Systems with
Service Air
Systems Used as
the Source ofBreathing Air

Date Issued
2/21792/2/792/91792/16/79

3/21179

3/23179
3/26/79

3/28179

Issued to
All power reactor
facilities withan OL and or CP
All fuel
facilities
All power reactor
facilities withan OL or a CP
All power reactor
facilities withan OL or a CP
All power reactor
facilities withan OL or CP
All holders ofreactor OLs or
CP
All power reactor
facilities with an
OL or CPAll power reactor
facilities withan OL and Pu
processing fuelfacilities

165

Information
Notice No.

Subject

Date Issued

Issued to

79-09

Spill ofRadioactively
Contaminated Resin

3/30179

All power reactor
facilities with
an OL

79-10

Nonconforming
Pipe Support
Struts4/16179

All power reactor
facilities with a
CP

79-11

Lower Reactor
Vessel Head
Insulation
Support Problem

517179All holders ofreactor OLs and
CPs

79-12

Attempted Damage
to New Fuel
Assemblies

5/11179

All fuel
facilities,
research reactors,and power reactors
with an OL or CP

79-13

Indication of Low
Water Level in
the Oyster Creek
Reactor
NRC Position on
Electrical Cable
Support Systems

5/29179

All holders ofreactor 0 Ls andCPs

6/11179

79-15

Deficient
Procedures

617179All power reactor
facilities with aCP
All holders ofreactor OLs and
CPs

79-16

Nuclear Incident
at Three Mile
IslandSource Holder
Assembly Damage
from Misfit
Between Assembly
and Reactor Upper
Grid Plate
Skylab Reentry

6/22179

All research
reactors and test
reactors with OLs

6/20/79

All holders ofreactor OLs andCPs

7/5179Pipe Cracks in Stagnant
Borated Water Systems At
PWR Plants
NRC Enforcement Policy
NRC LIcensed Individuals

7/17179

All holders ofreactor OLs
All holders of reactor
OLs and CPs

79-21

Transportation and Commercial
Burial of Radioactive Material

91717979-22

Qualification of Control
Systems

9/14179

79-23

Emergency Diesel Generator
Lube Oil Coolers

9/25179

79-24

Overpressurization ofContainment of a PWRPlant
After a Main Steam LineBreak

9/28179

79-14

79-17

79-18
79-19
79-20

8/10179

All holders of reactor
OLs and CPs and production
licensees with licensed
operators
All power and research
reactors with OLs
All power reactor
facilities withOLs and CPs
All power reactor
facilities holding OLs
and CPs
All power reactor facilities
with a CP

166

ed by a Bulletin, and a written response by the licensee
is not required. The licensees mayor may not initiate
the recommended action. However, if further analysis
and/or information regarding the matter indicates increased significance, it may result in the issuance of a
Bulletin.
The particular concerns which might require issuance of a Circular include those for which a Bulletin
is applicable, except that the impact is of lesssignificance and is not sufficient to warrant specificactions by license or permit holders.
The Information Notice was first put in use in 1979.
It is a mechanism by which the NRC is able to rapidlytransmit information applicable or potentially applicable to license and permit holders. The information mayor may not have been analyzed by NRC. It
does not require acknowledgment or response but
licensees are instructed to take appropriate action if
the information applies to their facility. The concerns
which might require issuance of an Information
Notice include those for which a Bulletin or Circular
may be applicable, but for which significance of theevent or condition does not warrant issuance of a
Bulletin or Circular. Of course, a Bulletin or Circular
may be issued subsequent to an Information Notice. ona particular concern asa result of problem evolutIOnand further evaluation. Information Notices may also
be used to transmit additional information on
previously issued Bulletins or Circulars to license andpermit holders.
A listing of the Bulletins, Circulars, and Information Notices issued from January 1, 1979, throughSeptember 30, 1979, is included in Table 3 to indicate
the types of conditions addressed by these different
publications.

Other Reactive Effort
During fiscal year 1979, the effort expended onreactive inspections, investigations and related work
has increased conSiderably, in addition to that expended on investigation and evaluation of the TMI accident.
Some construction sites have required between 50
and 250 man-days of unplanned reactive effort
resulting in some cases in the postponement of routine
inspection activities. A considerable amount of thisreactive effort relates to inspection, investigation andfollow-up effort, associated with allegations, Part 21
Reports and Bulletin, Circular and Information matters. The follOWing construction problems have required substantial reactive effort by both headquarters
and regional personnel:• Pipe support base plate/anchor bolts• Weld integrity (pipe welds)• Pump performance• Piping analysis and as-built conditions• Steam generators• Structural concrete• Foundations

ENFORCEMENT ACTIVITIESThe regulatory program is designed to assure thatlicensees perform in accordance with NRC regulations, licenses and permits and with applicable sections of Federal statutes. NRC is empowered to takeenforcement action when licensees are not satisfying
these requirements or are conducting operations in a
way that might endanger the public health and safetyor the environment, or adversely affect the common
defense and security.
Enforcement action may be taken, for example,
when certain significant safety-related matters notmeeting NRC requirements have escaped the licensee's
attention or when procedures are improperly controlled and the fact is first discovered during an NRC
inspection. Such situations reflect adversely on the effectiveness of the licensee's management or quality
assurance program. Enforcement action requires thelicensee to correct the particular problems andestablish measures to preclude reoccurrenceincluding deficiencies in his quality assurance program
if such deficiencies allowed the problem to occur, continue or reoccur.
The severity of NRC enforcement actions varies
with the seriousness of the matter and the licensee's
previous compliance record. Several levels of NRC action are provided:• Written Notices of Violation are provided for instances of noncompliance with NRC requirements.
• Civil penalties are considered for licenses who
evidence significant or repetitive items of noncompliance, particularly when a Notice of Violation has not been effective. 'Civil penalties may
also be imposed for particularly significant firstof-a-kind violations.
• Orders to "cease and desist" operations, or for
modification, suspension, or revocation oflicenses, are used to deal with licensees who do
not respond to civil penalties or to deal with violations that constitute a significant threat to publichealth and safety or to the common defense andsecurity. In the latter case, an order may be madeeffective immediately.
Tables 4 and 5 summarize the enforcement actions
taken during the report period.

167

Table 4. Civil Penalties Imposed-Fiscal Year 1979
LicenseeAmountReasonWisconsin Public Service
Corporation
Green Bay, Wisconsin(Kewaunee Plant)$7,000(reported as
Pending
in FY 78)

Failure to perform asurvey required byregulations to assure
control of personnel
exposures.
Licensee requested ahearing; however, anegotiated settlement
was accepted by thelicensee and thelicensee paid the$7,000 penalty.

Jersey Central Power andLight Company
Morristown, New Jersey(Oyster Creek Plant)$26,000

Twin City Testing andEngineering Labs., Inc.
St. Paul, Minnesota(Radiographer)$2,500

Failure to follow
radiation safety
procedures andnoncompliance itemsin the safeguards
area.
Exposure to the lower
back of an individual.
Failure to perform
necessary radiation
surveys.

Niagara Mohawk Power
Corporation(Nine Mile Point Unit 1)$18,000(pending)

Noncompliance itemsin the physicalsecurity area.

United Nuclear Corporation
Wood River Junction, RhodeIsland(Fuel Processor)$15,750(pending)

Noncompliance itemsin the physicalsecurity area.

University of Wisconsin
Madison, Wisconsin(Academic Broad
License)$2,300Inadequate trainingof personnel, failure
to evaluate internal
exposures of personnel
and releases of airborne
material to unrestricted
areas.

Virginia Electric and Power
Company(Surry Unit 2)$15,000(pending)

Nuclear Pharmacy, Inc.
Milwaukee, Wisconsin(Radiopharmaceu tical
Distributor)$24,000(pending)

University of Minnesota
Minneapolis, Minnesota$4,300Whole body exposureof an individual andfailure to follow
procedures.
Distribution ofradioactive material
not intended forhuman use to medical
licensees, relabelingand misrepresenting
the material as suitablefor human use.
Exposures of three
individuals to airborne
radioactive material andother noncompliance
items in the health andsafety area.

168

Enforcement Improvements
The Office of Inspection and Enforcement is seeking
continued improvement in enforcement. In December
1979 the enforcement criteria concerning the transportation of radioactive material were upgraded. The
Commission also has forwarded to Congress a request
to increase NRC's statutory authority to impose civil
penalties. If this request is implemented by amendment of the Atomic Energy Act, NRC's maximum
allowable penalties will increase from $5,000 to$100,000 for a single violation and from $25,000 to nolimit for all violations committed by a licensee within
30 days. Such an increase would provide greater incentives for major NRC licensees to comply with theregulatory requirements. A greater range would also
permit the penalties to be imposed by NRC to reflect
more equitably the different classes of licensees andthe seriousness of offenses. The Commission approved
a proposal that copies of escalated enforcement orders
and civil penalties be routinely forwarded to State
public utility regulatory groups and to State attorneys
general for their information. Routine mailing of thesecommunications started in December 1979.
NRC is continuing efforts to develop better methods
for the evaluation of the regulatory performance ofmajor lic~nsees. By identifying licensees whose performance may require improvement, NRC hopes to anticipate potential safety and security problems andavert them through prompt remedial action. This
would also improve the effectiveness of NRC's use of
inspection resources. Identifying valid measures oflicensee performance is a complex and controversial
process. Measures considered to date include licensees'compliance records, evaluations of licensees by NRC
inspectors, and detailed trend analysis of reportable
licensee events.

NRC Operations Center
The NRC Operations Center was activated on threeoccasions during 1979. This center is the focal point
for NRC's initial response to significant incidents involving NRC-licensed activities. The 2,000 square-footcenter presently in use includes: a conference room forbriefing NRC management; an operations room formonitoring and evaluating information about the incident; a secure communications room; word processing
and computer support areas; and a library to housenecessary information resources. The center is equipped with a specially-designed communications system
and a variety of audiovisual aids.
The first activation occurred in January as a resultof an extortion threat against the General Electric Fuel
Fabrication Facility in Wilmington, North Carolina.
A letter demanded money for return of stolen uraniumor the extortionist claimed he would disperse thematerial in an unnamed U.S. city. Although theFederal Bureau of Investigation (FBI) had the lead in
the case, the NRC was concerned about the possible
radiological consequences of the threatened act andprovided technical support to the FBI. In this case, theFBI quickly apprehended a suspect and located thestolen material.
The other major incident involved the NRC
response to the Three Mile Island accident. The centerdid function as a major focal point for the NRC, as intended, but the limited facilities were quickly overextended during this event. As a result of TMI and increased emphasis on responding to future incidents,
major revisions to the NRC incident response program
will be made.
The third incident for which the Operations Center
was activated occurred in October when a release ofradioactive gases from the Prairie Island nuclear plant
took place.
The Operations Center is manned 24 hours-per-day
by a qualified senior engineer.

INVESTIGATIONSAn important adjunct to NRC's inspection effort is
the investigative program which covers not only indepth probes of irregularities revealed during inspections, but also investigations of incidents, accidents,
allegations or any unsual circumstances occurring ator related to NRC-licensed facilities or activities. A
heightened public awareness and interest in nuclear
power has resulted in an increase in the number ofallegations received by NRC. As each allegation must
be carefully investigated to determine its possible impact upon the public health and safety, NRC has morethan doubled the number of trained investigators in its
employ within the past year.
Investigations are conducted by experienced investigative personnel located in each of the five NRC
regional offices. Investigators are assigned to the immediate staff of the regional director, both to emphasize the importance of the investigative program
and to provide better support to the various functional
branches in the region. Since NRC investigations areusually technical in nature and may involve several
scientific or engineering disciplines, the investigatorfrequently works with and coordinates the activities oftechnical personnel who may be assigned to provide
assistance. Investigators also maintain close liaison
with Federal, State and local law enforcement agencies and work closely with them on investigations ofmutual interest. Within the past year, IE investigators
have provided assistance to agencies having primary
jurisdiction in investigations involving the theft ofspecial nuclear material, the intentional damaging offuel elements at an operating nuclear power plant, theattempted bombing of a nuclear power station, andthe falsification of records relied upon by NRC.

169

Table 5. Enforcement Orders-Fiscal Year 1979
LicenseDate

ReasonRadioassay Systems, Inc.
Southfield, Michigan(Materials Licensee)

11130178

Order terminatingproceedings.
Reason: Licensee
disposed of allmaterial and requested
termination of thelicensee. On 7/13178,
the licensee was issued
an Order to show cause
for processing and distributing
material without
au thoriza tion.

Arkansas Power & Light
Company
Little Rock, Arkansas(Arkansas Nuclear One
Unit 1)

6/15179

Public Service of Indiana
Plainfield, Indiana(Marble HillU nits 1 & 2)

8/15179

Order authorizing
resumption ofoperations.
Reason: Licensee
satisfied theconditions of the
6/2179 Order.

Oversight of the NRC investigations program is accomplished by a small investigative staff located at
headquarters. During fiscal year 1979, 121 investigations were conducted by inspection and enforcement
personnel. Of these, 76 were prompted by allegations
dealing with reactor construction or operational
events at licensed facilities. Other investigations wereconducted into events involving loss or theft of licensed
material, overexposures, and general public interest.
In 78 of the investigations, licensees were cited for
failure to meet NRC requirements.
Significant special investigations conducted during
the year are described below.

Wolf Creek Generating Station
The Wolf Creek Generating Station of the Kansas
Gas and Electric Company is located in east-central
Kansas in Coffey County. The site is approximately 50

Order confirming
suspension of
construction.
Reason: Serious
problems with respect
to the adequacy of
concrete placement
and the licensee'squality assurance
program.

miles south of Topeka and three miles northeast of
Burlington, Kansas.
On March 15, 1978, the licensee reported to the
NRC that concrete samples tested for compressive
strength at the age of 90 days had not all met the
specified 5,000 Ibs.-per-square-inch (psi) design
strength. The samples represented 6,600 cubic yards ofconcrete placed in a continuous two-day operation to
construct the reactor containment building base mat.
The licensee initiated a series of studies to determine
the cause of the low strength in samples and to determine whether the base mat met the construction permit criteria. The licensee concluded, in a final reportin October 1978, that the base mat was acceptable andmet specifications for 5,000 psi compressive strength
concrete on the basis of supplemental tests. The apparent low-strength samples were attributed to faulty
testing procedures by the licensee.

170

The NRC initiated an investigation that resulted in
the licensee's halting the placement of all safetyrelated concrete on December 19, 1978. Numerous
deficiencies which could have contributed to the apparent low-strength concrete were found, as well as
quality assurance problems. The NRC concluded thatthe faulty testing was not the cause of the apparent
low-strength samples. Additional studies were initiated by the licensee. The NRC retained an independent consultant and a test laboratory to provide additional information independent of the licensee.
While this work was underway, voids were found in
the containment wall in two locations when forms
were removed in December 1978. These defects andtheir causes were reviewed by IE inspectors. Repair
was subsequently accomplished, utilizing approved
procedures.
On March 6, 1979, after changes related to quality
assurance had been made, the licensee was permitted
to resume the placement of concrete in safety-related
structures except for the reactor containment building.
It was not until July 12, 1979 that concrete placement
was permitted in the containment, because of theunresolved questions concerning the base mat. Placement was allowed as a result of reanalyses of the reactor containment building by the licensee using thelowered concrete strength values as the actual as-built
strength of concrete. The reanalysis showed thatenough margin remained in the design to accommodate the low-strength concrete, since the Wolf
Creek unit is one of a series of standardized plants
which are designed for more severe site conditions
than exist at Wolf Creek. As a result of the studies andinvestigations, greater assurance has been obtainedthat the structure will perform adequately.

Marble Hill Nuclear Generating Station
The Marble Hill Nuclear Generating Station of
Public Service of Indiana is located in southeastern Indiana in Jefferson County. The site is approximatelynine miles northeast of Milton, Ky.
Beginning in April 1979, a series of noncompliances
associated with concrete construction were identified
by IE inspectors. In May, the NRC met with thelicensee to request additional information on in-place
concrete. Many of the noncompliances were attributed
to inadequate implementation of quality
assurance/ control programs by the licensee and his
contractor. In June, a series of allegations related toconcrete construction were made by a former worker
at the site. These allegations indicated that voids in theconcrete had been found but not properly reported nor
properly repaired.
An NRC investigation, with the aid of the worker,
found additional areas which were deficient becauseof voids. In June, the licensee agreed to stop safety-

related concrete work until certain QA actions werecompleted to the NRC's satisfaction. On the basis ofobservation by IE inspectors of a large non-safetyrelated concrete placement, safety-related concrete
work was allowed to resume.
In July 1979, the National Board of Boiler andPressure Vessel Inspectors reported several deficiencies
at the site (not related to concrete) and recommended
suspension of the utility's American Society of
Mechanical Engineers Owner's Certificate for apparent Code violations of Section III, Division 1, of
the ASME Boiler and Pressure Vessel Code. As a resultof further investigation by IE inspectors, which identified contruction management problems, an orderconfirming the suspension of work was issued on
August 14, 1979. A series of corrective actions must becompleted before construction of safety-related items
will be allowed to resume.

McGuire Nuclear Plant
The McGuire Nuclear Plant of Duke Power Company is located 17 miles northwest of Charlotte, N. C.,
adjacent to the Catawba River.
In March 1978, the NRC received telephone calls
and a letter from an individual regarding alleged safety problems at the McGuire facility. A meeting of
NRC staff with the individual resulted in reduction of
the concerns to 12 allegations. IE investigators workedon the case through July and were able to resolve all
but one allegation. This one allegation pertained to
calculations completed by Duke Power Company toascertain whether a fuel cask could fall into the spent
fuel pool under various hypothetical circumstances.
Additional investigatory effort identified a conservative calculation-not previously shown to the
NRC-which showed that the cask could enter the
spent fuel pool. The license has taken corrective action
to prevent the occurrence of such an event.

Midland Nuclear Plant
The Midland Nuclear Plant, which is owned by
Consumers Power Co., is located just south ofMidland, Michigan, adjacent to a large industrial
complex of the Dow Chemical Co.
In September 1978 the licensee reported greaterthan expected settlements had occurred in the diesel
generator building complex. IE investigations disclosed that many of the commitments the licensee had
made at the construction permit stage had been revised without changes in the safety analysis reports. Matters related to revised criteria and remedial action
were transferred to NRC's Office of Nuclear Reactor
Regulation (NRR).
Joint IE-NRR efforts are still underway to define
what corrective measures need to be taken.

=================================================171
Surry Nuclear Power Station
The Surry Nuclear Power Station, which is ownedby the Virginia Electric and Power Company (VEPCO), is located about eight miles south ofWilliamsburg, Va.
On May 7, 1979, while conducting inspections of
new fuel for Surry Unit 2, the licensee found thatplastic protective liners on 62 of 64 nuclear fuelassemblies had been tampered with. Further inspection revealed that a white crystalline substance had
been poured on the assemblies. Preliminary analysis
by VEPCO indicated that the substance was sodiumhydroxide. The new fuel is stored in a building which
is locked and alarmed, and to which access is controlled by the issuance of specially coded access cards.
Investigation of this incident by the FBI culminatedin the surrender of two VEPCO employees to SurryCounty authorities on June 19, 1979. Charges ofbreaking and entering with intent to damage electricalfacilities (felony) and willful destruction of utility
company equipment (felony) were lodged against thetwo employees. Two additional charges against both
men were introduced for conspiracy regarding the two
felonies, for a total of four felonies and one misdemeanor against each man. These charges were filed onbehalf of the Commonwealth of Virginia, since current Federal statutes do not provide penalties for suchacts of vandalism. Pending legislation may result inmaking intentional damage to a reactor facility aFederal crime.
The two individuals, who were later tried, convicted on charges of willfully destroying utility com·pany equipment and sentenced to two years' imprisonment, claimed that they had damaged the fuel rods tocall attention to poor security practices and unsafe
conditions at the VEPCO facility. Subsequent to theirtrial, they were interviewed by NRC investigators andan investigation into their allegations is currently
underway.

Abnormal Occurrences-Fiscal Year 1979
An "abnormal occurrence" is defined in Section 208of the Energy Reorganization Act of 1974 as "anunscheduled incident or event which the Nuclear
Regulatory Commission determines to be significant
from the standpoint of public health or safety." Thesame Act requires that such events be' reported
quarterly to the Congress by the NRC and also be included in the Annual Report. The four quarterly
reports covering fiscal year 1979 are published as
NUREG-0900, Vol. 1, No.4, and Vol. 2, Nos. 1,2 and3, and are available from the Division of Technical Information and Document Control, U.S. Nuclear
Regulatory Commission, Washington, D.C. 20555
and from the National Technical Information Service,
Springfield, Va. 22161.

In reports on the first three quarters of the fiscal
year, eight abnormal occurrences were covered, including the accident at Three Mile Island (see Chapter2). Three additional events were under consideration
for reporting in the fourth quarter (July-September1979) but had not been officially identified as abnormal occurrences at the end of the report period andare, therefore, omitted in the listing below. Abnormal
occurrences which took place during the fiscal year at
facilities under the jurisdiction of Agreement States
are treated in Chapter 8, "State Programs." .
Loss of Containment Integrity. The occurrence involved a loss of containment integrity at two nuclear
power plants-Mil1stone Unit 2 and Salem Unit
I-reported in July and September of 1978 respectively. The issue is discussed in Chapter 3, under"Mechanical Operability' of Containment Purge
Valves," in the section, "Other Technical Issues."
Electrical System Deficiencies. The occurrence concerned degraded engineered safety features at the
Arkansas Nuclear One site, involving both Units 1 and
2 there, and disclosed serious deficiencies in electrical
distribution system operation and design. It wasreported in September 1978 and is treated in
NUREG.0090, Vol. 2., No.1
Piping Reanalysis at Five Plants. The occurrence
derived from the discovery that certain piping systems
and pipe suports in five nuclear plants had been constructed according to a faulty calculation. The issue iscovered in Chapter 3, under "Shutdown and Seismic
Reanalysis of Five Operating Reactors," in the section,"Other Technical Issues."
Extortion Attempt. The occurrence arose from an
extortion attempt in the form of an anonymous letter
sent to officials of the General Electric Company's fuelfabrication facility at Wilmington, N.C., alleging thatthe sender was in possession of an amount of lowenriched uranium oxide and threatening to send portions of it to various persons and to release the materalin certain cities if payment was not made. The extortionist was apprehended and the material recovered.
(See Chapter 5, under "Safeguards Events-Fiscal
Year 1979.)
Loss of Feedwater Transient. The occurrence took
place on May 2, 1979, at the Oyster Creek facility,
where a loss of feedwater transient resulted in a signifi.
cant reduction of the water inventory above the reac-·
tor core area, as measured by one set of water level instruments. It was later determined that the water level
had fallen below the safety limit, but that no part ofthe core was uncovered and no fuel damage occurred.
Vandalizing New Fuel Rods. The occurrence involved the pouring of sodium hydroxide on new fuelassemblies and is discussed above under the heading"Surry Nuclear Power Station."

172

Emergency Feedwater Unavailable. In June 1979 at
Unit 1 of Arkansas Nuclear One, an NRC inspector
found that, as preparations were made for startup of
the facility the controls for the emergency feedwater
system were so positioned that the system could not
automatically respond if needed. It was later ascer-

tained that there was no procedural requirement thatthe system status be checked before startup. The plant
was returned to cold shutdown for 12 days, until procedures could be reexamined and revised. All holdersof reactor operating licenses and construction permits
were informed of the event and its implications.

8State Programs
NRC and State officials dealt with milltailings matters throughout 1979.

The Three Mile Island nuclear power plant accident
in Pennsylvania focused increasing interest by theStates on most areas of NRC activity, including reactor
regulation, emergency preparedness, and waste
disposal. While NRC's contacts with the States are far
ranging and involve activities of many of the agency's
offices as well as the Commission itself, the principal
responsibility for NRC/State interaction is centered in
the Office of State Programs.
Highlights of fiscal year 1979 included the negotia~
tion of memoranda of understanding with Indiana
and Nebraska, regional meetings with State liaison officers in NRC Regions I and II, a decision to place
State liaison officers in all NRC regions, NRC concurrence in five more State plans for response toradiological emergencies, and several regional
workshops to develop a more explicit policy for
nuclear facility decommissioning.
NRC/State activities discussed in this chapter include (a) the State Agreements Program, under which
NRC relinquishes to qualified States the authority toregulate certain kinds and quantities of nuclear
materials; (b) assistance to State and local governments in radiological emergency response planning;and (c) cooperative activities regarding NRC responsibilities affecting the States such as licensing, decommissioning, waste management, and transportation ofradioactive materials.

STATE AGREEMENTS PROGRAM
Section 274 of the Atomic Energy Act of 1954, as
amended, authorizes the Commission to enter into
agreements providing for the assumption by qualified
States of regulatory responsibility over byproduct andsource material and small quantities of special nuclear
material. At the end of 1979, there were 25 Agreement

States exerClsmg regulatory authority over some
11,800 nuclear material licenses: Alabama, Arizona,
Arkansas, California, Colorado, Florida, Georgia,
Idaho, Kansas, Kentucky, Louisiana, Maryland,
Mississippi, Nebraska, Nevada, New Hampshire, New
Mexico, New York, North Carolina, North Dakota,
Oregon, South Carolina, Tennessee, Texas, andWashington. (An agreement concluded in late 1979
with the State of Rhode Island became effective on
January 1, 1980.)

Review of State Regulatory Programs
The NRC conducts a formal annual review of each
Agreement State's radiation control program to determine whether it is adequate to protect the publichealth and safety and is compatible with NRC's
regulatory program. The annual reviews assess theState's organization, administration, staffing, regulations, licensing, and compliance functions for the pro~gram. Field evaluations of State inspectors are also
made. During fiscal year 1979, the NRC conducted 29
such program reviews and one followup review. NRC
staff accompanied State inspectors at a number oflicensed facilities, including four State-licensed
uranium mills.

Adequacy and Compatibility Findings
During calendar year 1978, NRC found that all 25
Agreement State radiation control programs were adequate to protect public health and safety. The NRC
staff did, however, recommend a follow-up review of
the Florida program because of a recurring high inspection backlog and staff shortages.
With respect to the compatibility of Agreement
State programs with NRC's regulatory program, NRC
determined that 23 of the 25 States had compatible
programs in calendar year 1978; however, the pro-

174

NRC staff members meet periodically with representatives of the
National Association of State Directors for Disaster Preparedness,
the U.S. Civil Defense Council, and the Conference of State

Radiation Control to seek their views and assistance in developing
and improving Federal programs for radiolOgical emergency
response preparedness.

grams of Nevada and New Mexico were found not fully compatible. Compatibility findings for these two
States were deferred because they had not adopted
regulations fully equivalent to those of the NRC dealing with requirements for notices, instructions, andreports by licensees to workers (10 CFR Part 19 of
NRC regulations).

turer of devices containing tritium. Texas, New Mexico, Colorado, Washington, and Arizona are receiving
NRC assistance connected with uranium milling
operations, and NRC is giving a great deal of technicalassistance to Arizona in a case involving excessive
radioactivity released from a plant manufacturing
devices containing tritium.

Training Offered by NRC
NR C Technical Assistance
NRC provides technical assistance to the Agreement
States in areas such as major licensing actions, healthphysics, environmental analyses, review of proposed
regulations, and guidance for inspection and enforcement actions. NRC is assisting Kansas with the review
of a proposed low-level waste repository, andWashington in connection with renewal of a low-level
waste disposal license. Nevada asked for and received
NRC assistance regarding waste shipments to theburial site at Beatty. New York received NRC help in
its review of an environmental report from a manufac-

State regulatory personnel regularly attend NRCsponsored courses to upgrade technical and administrative skills. This training is available to both
Agreement and non-Agreement State personnel at nocost.
The following training courses were presented infiscal year 1979: Safety Aspects of Industrial
Radiography at Louisiana State University; Medical
Uses of Radionuclides at Baylor College of Medicine in
Texas; Health Physics and Radiation Protection at Oak
Ridge Associated Universities; Inspection Proceduresat NRC Region III Offices; Calibration of Teletherapy
Machines at M. D. Anderson Hospital; and Orienta-

==========================================~======175tion in Regulatory Practices and Procedures at NRC
Headquarters. In all, 138 State personnel received 292
student-weeks of training during the year.

Annual Meeting
The Agreement State radiation control program
directors meet each year at NRC headquarters to
discuss a wide range of topics. The October 1978 agenda covered the transportation of radioactive materials,
high- and low-level waste, generally licensed devices,
occupational ALARA (maintaining a level of
employees' exposure to radioactivity which is "as low
as reasonably achievable"), decommissioning, offshore
radiography, and uranium mills. The meeting produced recommendations from State representatives to the
NRC about the uniform Federal regulation of all
naturally occurring and accelerator-produced
radioactive material, as well as "agreement" material;
the registration of industrial radiographers; resolutionof the nation's waste disposal problem; and NRC's
regulations pertaining to transportation.

for hearings and public participation in the processingof license applications for these facilities. The Agreement States regulate more than half of the active
uranium mills and have a number of abandoned tailings piles within their borders. In conformance withthe legislation, NRC will be negotiating amendments
to existing agreements with those States that currently
license uranium mills and wish to continue such
regulation. These agreements, plus technical andfinancial support in the form of grants, will result inuniform regulation of uranium mills across the nation.
(See also Chapter 6.)

Agreement State Abnormal Occurrences

The Uranium Mill Tailings Radiation Control Act of1978 requires Agreement States which wish to continue regulating uranium mills and tailings afterNovember 8, 1981, to adopt Federal technical standards and procedures, including the preparation of
written analyses, and the provision of opportunities

In early 1977 the Commission directed that abnormal occurrences taking place at facilities of Agreement
State licensees should be included in the quarterly
report to the Congress (see Chapter 7, "Abnormal
Occurrences-Fiscal Year 1979"). The criteria appliedin determining that an event at an Agreement State
licensee's facility is an abnormal occurrence are thesame as those applied to NRC licensees.
During the first three quarters of fiscal year 1979
(October 1978 through June 1979), a total of four abnormal occurrences were reported to Congress
which took place in Agreement States. Two additional
events were under consideration for reporting in the
fourth quarter but had not been officially identified asabnormal occurrences at the end of the report period.

NRC staff and State representatives exchange information andideas at the Annual Agreement States meeting held in Silver
Spring, Md., in Octol)er 1979. The discussion included issues on

emergency response, transportation and waste management,
regUlation of uranium mills and environmental review, and sitingof nuclear power plants.

Agreement States and Mill Tailings

176

Two of the occurrences reported from Agreement
States took place in the first quarter of fiscal year 1979
and are covered in the quarterly report to Congress,
NUREG-0900, Vol. 1 No.4. One of these involved theoverexposure of a radiographer's assistant in Louisiana
and the other the transportation of a package ofradioactive material whose radiation emission after
packaging exceeded limits set out in the license of thesender.
During the third quarter (covered in NUREG-0900,
Vol. 2, No.2), two more events were reported from
Agreement States as abnormal occurrences. On March
9, 1979, the Arizona Atomic Energy ICommission
found several items of noncompliance in the operations of a State licensee engaged in making anddistributing to authorized persons various signs anddevices using tritium as an activating agent. An unannounced inspection on May 7 revealed continued noncompliance and also the presence of tritium in foodprepared in a facility near the licensee for a number ofschools in the area. The level of tritium exceeded the
EPA standard for tritium concentration in liquids by
180 percent. The company was directed by the State todecommission operations, the tritium on the premises
was sealed up and, by order of the Governor, removed
to a U.S. Army facility leased for the purpose.
In California, a State licensee was conducting
radiography activities at a manufacturing plant on
May 22, 1979. The radiographer failed to notice thatthe radioactive source in his instrument had become
disconnected. It was found on the floor by a plant
employee who put it in his hip pocket. He later passed
it on to another employee of the plant and a number ofothers also handled the source before it was retrievedby the radiographer. The radiographer did not informthe nine people who had been exposed to the source of
its radioactivity and the attendant dangers, nor did hereport the incident to either his own or the client's
management. The employee who had picked up andpocketed the source was later hospitalized and required surgical repair of ulcerated skin. It is estimatedthat he had received a dose on the skin surface of 1.5
million rem. Others exposed to the source received
radiation doses in the thousands of rem to their hands,
and several incurred radiation burns. The Statesuspended the radiography firm's license and instituted a State Board of Inquiry to investigate thematter. NRC alerted all radiography licensees to theevent and to the importance of the training ofradiographers, of their performing radiation surveys,
and of their promptly notifying responsible management in the event of accidental exposures to radiation.

EMERGENCY PREPAREDNESSEmergency Response Planning
The accident at Three Mile Island (TMI) has greatly
intensified interest in emergency preparedness on thepart of the public, the Congress, and the NRC. In thepast, State and local government efforts in this field
have been largely voluntary. There was no requirement that States do such planning either by law or by
rule, and no sanction could be visited upon a State or
locality which chose to neglect or ignore the subject.
In the wake of TMI, there has been widespreadrecognition that too little attention had been paid toemergency preparedness in the past and that much
more time, effort and money mlist be devoted to it in
the future by NRC, other Federal agencies, State andlocal governments, and the nuclear utilities. In· thefuture, the present voluntary system for reviewing
State and local plans may well, and probably will, be
replaced by a more formal system, based on legislation
or regulations, or both.
The Procedure. The responsibilities of Federal agen.
cies for assisting State and local governments indeveloping plans for responding to radiologicalemergencies were outlined in a Federal Register noticeof December 24, 1975, promulgated by the former
Federal Preparedness Agency (FPA) of the General
Services Administration. The notice, entitled"Radiological Incident Emergency Response Planning; Fixed Facilities and Transportation," gave the"lead agency" role to NRC, while assigning specificsupport responsibilities to the Environmental Protection Agency (EPA); the Department of Energy (DOE);
the Department of Transportation (DOT); theDepartment of Health, Education and Welfare(HEW); the Defense Civil Preparedness Agency(DCPA); and the Federal Disaster Assistance Administration (FDAA) of the Department of Housing
and Urban Development. Under powers granted him
by the Congress, President Carter combined three ofthese agencies (FP A, DCPA and FDAA) into a newFederal Emergency Management Agency (FEMA) on
July 15, 1979.
In his statement of December 7, 1979, responding to
the report of the President's Commission on the Accident at Three Mile Island, President Carter directedthat FEMA: "(1) take the lead in off-site emergency
planning and response; (2) complete by June 1980 thereview of State emergency plans in those states with
operating reactors; (3) complete as soon as possible thereview of state emergency plans in those states withplants scheduled for operation in the near future; (4)develop and issue an updated series of interagency
assignments which would delineate respective agency
capabilities and responsibilities and clearly define
procedures for coordination and direction for both
emergency planning and response; (5) assure that
DOE resources and capabilities for responding toradiological emergencies are made available andaugmented as needed to service civilian related
radiological emergencies; and (6) assure the development of programs to address the recommendations for
additional research and public education needs."
NRC is cooperating fully with all of these efforts of
the new agency (see Chapter 1 and Chapter 2).
Concurrence in State Plans. Six State plans received
NRC concurrence in 1979, bringing to 14 the numberof State plans so approved.

===================================================177
Planning Guidance to States
NRC has been working with the EPA to determine
the types of accidents for which radiological emergency plans should be developed by State and local
governments. A draft report on this subject (NUREG0396/EPA 520/1-78-016) was completed by the
NRC/EPA Task Force on Emergency Planning andissued for public comment in December 1978. Tne task
force concluded there was no specific accident sequence that could be used for emergency planningoecause each accident could have different consequences, both in nature and degree. Instead, the task
force developed recommendations in an alternative
form which would provide State and local governments with a basis on which to formulate emergenc),
plans. The planning basis selected involves a variety ofaccident consequences. The planning distances, time
characteristics, and radiological release characteristics
specified in the report provide guidance that scopes
the emergency planning effort.
The fundamental recommendation in the NRC/EPA
task force report is that Emergency Planning Zones(EPZs) be established around each nuclear power
plant for purposes of emergency planning, and that an
EPZ of about 10 miles in radius be established for theplume exposure pathway and a second concentric EPZof about 50 miles in radius be established for the ingestion exposure pathway (milk and agricultural products).
The final report was published for public comment
on December 15, 1978. The original 90-day comment
period was extended to May 15, 1979 as a result of theThree Mile Island accident. The task force recommendations were submitted to the Commission in July
1979, and Commission action is expected early in fiscal
year 1980.

Training Program for States
Several years ago, in cooperation with the Statesand other Federal agencies, NRC identified a numberof areas where training was needed for State and local
government personnel involved in radiologicalemergency planning and preparedness. Three training
courses are now being offered. Courses dealing withradioactive materials in transit will be developed by
DOT during fiscal year 1980, and courses in the
medical area are being considered. FEMA is planningcourses for "first-at-the-scene" personnel.
The following training is offered free of charge toqualified State and local government personnel:(1) Radiological Emergency Response Operations:This course is now conducted routinely at
DOE's Nevada Test Site. It is deSigned for personnel who are, or will be, assigned to State or
local radiological emergency response teams.
Sixteen sessions were conducted during fiscal
year 1979 for 320 State and local government
employees. Eighty Federal employees received
training in the same program.

At DOE's Nevada Test Site, NRC sponsors training in radiolOgicalemergency response operations for State and local govemment personnel who are or may be members of response teams duringemer~encies. Above, students conduct a survey of contamination
resulting from a simulated ground spill, while a faculty member
acts as a news correspondent. Below, students "suit up" before
entering a contaminated area.

178(2) Radiological Emergency Response Coordination: This course is designed to help the State
radiological emergency respons~ coor~inatormake decisions on what protective actions totake in the event of an accidental release ofradioactive material to the environment from anuclear facility. The course is conducted on request by the States.
(3) Radiological· Emergency Response Planning:This course was developed to provide training
needed for State and local radiologicalemergency response planners, and is conducted
on request.
(4) Handling Radioactive Material in Transportation Accidents: Through the interagency program described in t.he Dece~ber 24, 1~75
Federal Register notIce, and In cooperatIOn
with NRC, the DOT developed an 8-hour training course on handling radioactive m~terial intransportation accidents. The course IS a selfcontained package consisting of slides andtaped narratives and a student workbook. One
package will be made available free of charge
to all States by DOT, and NRC and DOT plan
to make it available to many local jurisdictions.

Field Assistance Program
NRC continues to lead and coordinate Federal interagency field reviews of State radiological emergency response plans and critiques of exercis~ to test ~heseplans. During fiscal year 197~, the regIO~al advI~o~ycommittees made 35 field reVIew and asSIstance VISItS
and critiqued 12 radiological emergency response exercises.

TMI Activities
Like many offices within NRC, the Office of State
Programs' staff spent considerable time on Three Mile
Island (TMI) activities and subsequent followups. In
the early stages of the TMI accident, six healthphysicists from the Agreement States Program went tothe site to assist in a variety of tasks, including environmental sampling, communications, and direct
health physics technical support to the State of Pennsylvania. This entire NRC activity is covered in
Chapter 2 and in other reports. It is important to note
that, as a result of the accident, many States which
previously were not actively pursuing concurrence intheir radiological emergency response plans are now
actively seeking such concurrence. Many meetings
were held with States; office personnel testified atseveral State and Congressional hearings on the subject; and plans and schedules were made to concur inplans of 18 additional States by May 1980. To help

with this new workload, personnel were temporarily
assigned to the Office of State Programs from the Office of Nuclear Reactor Regulation, and temporary
employees and consultants were acquired.

GAO Report
The General Accounting Office (GAO) published areport March 30, 1979, entitled "Areas Around
Nuclear Facilities Should be Better Prepared forRadiological Emergencies." The report made recommendations to the Secretaries of Defense and Energy,
the Director of the Federal Emergency Management
Agency, and the Chairman of the NRC.
The GAO recommended that no nuclear power
plant be allowed to begin operations until State andlocal emergency response plans contain all the Commission's essential planning elements, and thatlicensees make arrangements for State and local agency participation in annual emergency drills. The Commission responded that NRC is committed to having
effective, tested emergency plans wherever needed
and as early as possible, and that the proposed licensing requirement dealing with plans and exercises will
be included in an expedited NRC rulemaking procedure.
The GAO recommended that NRC establish the
10-mile emergency planning zone around all nuclear
power plants. The Commission has endorsed this concept, as previously mentioned.
The GAO recommended that there be a requirement for people living near nuclear facilities to begiven information about the potential hazard, theemergency actions planned, and the proper course ofaction in case of a radiological release. The Commission response said that action will be taken to implement this recommendation in connection with NRC's
ongoing assessment of regulatory requirements.

Other Emergency Response Activities(1) Under a contract with DOE, Sandia
Laboratories is developing a set of accident
scenarios which can be used to test nuclear
facility, State and local government emergency
plans.
(2) To answer the need for improved emergency
planning guidance in the event of transp~rta­tion accidents involving radioactive matenals,
an NRC/DOT task force will be established inearly 1980 to deal with the subject.
(3) A large step was taken in 1979 to provid~ mo~e
uniformity in reviewing and concurrmg m
State/local plans. At a national meeting ofFederal regional personnel involved in the179review process, acceptance criteria weredeveloped for each of the essential elements required for concurrence. These criteria will be
used to judge the adequacy of individual
elements. Such a system eliminates much of thesubjectivity involved in differing interpretations of what constitutes acceptability. The
criteria are intended for use by both planners
and reviewers.
(4) A draft report called "Beyond Defense In
Depth" (NUREG-0553) was published in
March 1979. It is a study of the costs of developing and implementing State and local emergency response plans, which are particularly acuteat the local government level. It also discusses
several methods of funding such plans andrecommends that additional funds for emergency planning by State and local governments beraised through the imposition of additional feeson licensees and on applicants for NRC licenses.
The final report will be published for publiccomment. The NRC staff plans to make formal
recommendations to the Commission and to the
new FEMA concerning the funding problem
and possible solutions to it.

LIAISON AND COOPERATIVEACTIVITIESTransportation Surveillance
During fiscal year 1979, seven States participated in
the NRC/DOT program for the surveillance ofradioactive material transported into, within or
through their borders. Georgia, Illinois, Michigan,
and South Carolina completed 2 years of monitoring.
The first-year results of the Illinois program (for theperiod June 1977 to June 1978) and the Georgia program (August 1977 to September 1978) were published
as NUREG/CR-0756 and .0931, respectively. Kentucky will complete its first year of monitoring in
December 1979. Washington and Florida began theirprograms in September.
The program contributes valuable data concerning
all aspects of transportation in the respective States;
promotes greater familiarity with Federal and Stateregulations on the part of shippers, carriers, and State
personnel; and results in closer adherence to theregulation, thus safeguarding the health and safety oftransportation workers and the general public.

Memorandums of Agreement
In January 1976, NRC and EPA entered into a second memorandum of understanding regarding theirrt:spective responsibilities under the Federal Water

An NRC radiation specialist checks a trailer carrying low-level
radioactive waste materials. Shipments such as this one from Three
Mile Island are checked frequently to ensure that radiation iswithin safe limits.

Pollution Control Act Amendments of 1972 (FWPCA).
NRC encourages agreements with States to whom EPA
has delegated the National Pollutant Discharge
Elimination System (NPDES) permitting authority
under section 402 of the FWPCA.
In the recent past, NRC entered into understandings
with Virginia, New York, South Carolina andWashington. During fiscal year 1979, NRC concluded
memorandums of understanding with Indiana and
Nebraska. Discussions continue with several otherStates.

State Liaison Officers Program
The Governors of all States have appointed liaison
officers to maintain direct communication with NRC.
There are now a total of 51 State liaison officers to the
NRC, from the 50 States as well as the Commonwealthof Puerto Rico.
Regional State Liaison Officers' meetings were heldin NRC Region I in October 1978 in King of Prussia,

180

Pennsylvania, and in Region II in December 1978 in
Atlanta, to acquaint the States with the NRC regional
office operations and to discuss such major issues as thetransportation of radioactive waste, decommissioning,
waste management, emergency planning and notification. Meetings are also planned for NRC Regions IV
and V during the fiscal year 1980.
A pilot program by NRC to place its own Stateliaison officers in the Philadelphia and San Francisco
regional offices was begun by the Commission in 1977.
After evaluating the program in July 1979, NRC
approved full implementation of this program to all
regions. During fiscal year 1980, NRC expects toassign liaison officers to the other three regions.

meeting was the sharing of Federal and State experiences as a result of the Three Mile Island accident.
The Interorganizational Committee for Radiological Emergency Response Planning and Preparedness-comprised of representatives of the Conferenceof Radiation Control Program Directors, the National
Association of State Directors for Disaster Preparedness, and the U.S. Civil Defense Council-continued
to review Federal guidance publications, training,
field assistance and other Federal emergency planningand preparedness efforts, and NRC will continue tolook to this group to provide the State and local
government comments and suggestions.

Intergovernmental Personnel Assignments
National/State Organizations
Throughout 1979, NRC engaged in cooperative efforts with regional bodies such as the Western Interstate Energy Board, and with national State
organizations such as the National Governors' Association, National Conference of State Legislatures, National Association of Attorneys General, National
Association of Counties, and National Association ofRegulatory Utility Commissioners. NRC staff also met
with State legislators several times during the year to
discuss NRC's programs on radioactive waste management, decommissioning of nuclear facilities andradiological emergency response planning.
There was a great deal of legislation concerned with
nuclear power before State legislatures during the1979 legislative session. Numerous bills dealt withradioactive waste, transportation, emergency response
planning, and conditioning nuclear power plant siting
to solution of the waste problems. NRC continued toprovide comments on proposed legislation when requested and in several instances presented testimony
before legislative committees.

During 1979 NRC professional employees were serving with the States of Arizona, Georgia, New Mexicoand Oregon under the provisions of the Intergovernmental Personnel Act of 1970. Under the same program, a staff member of the New Mexico Environmental Improvement Division was detailed to the NRC.

Conference of Radiation Control Program
Directors
Together with the Food and Drug Administration
and Bureau of Radiological Health of HEW, and the
Office of Radiation Programs of the Environmental
Protec-tion Agency, the NRC continued its financial
and technical assistance to the Conference of Radiation Control Program Directors, Inc. The Conference,composed of the heads of State and major municipal
radiological health programs, seeks to promote andcoordinate State and local radiological health activities. A major topic of the Conference's 1979

The first shipment of low-level radioactive waste from the damaged Three Mile Island Unit 2 arrives at the Nuclear Engineering
Company's (NECO) Hanford, Washington, disposal site. Inspecting the load are Beth A. Riedlinger, a radiation specialist from
NRC region V office (on step ladder), David Jenkins, Special Assistant to the Governor of Washington (right), and a NECO
employee.

===================================================181
NRC served as host agency for the civil defense
director of Westchester County, New York, as part of
the Intergovernmental Affairs Fellowship Program.
He was assigned by the Office of Personnel Management to the NRC Office of State Programs to help improve radiological emergency response planningguidance for county governments and to assist in setting up an integrated emergency response plan for thefive counties surrounding the Indian Point nuclear
power facility in the State of New York.

the 1978 workshops, (2) a preliminary draft of ageneric environmental impact statement, and (3) draft
proposed rule changes, were held in September 1979
in Columbia, South Carolina, and in Seattle,
Washington. These workshops were attended by abroad cross-section of State officials, including
legislators as well as energy policy, siting, economic,
regulatory, and radiation control officials. More than
150 State officials from 44 States participated.

Workshops
NRC has found State personnel a valuable source ofhelp in the development and, more particularly, thereview of policy and regulations. Experience has
shown that the regional workshop is a most effective
tool in this process.
For some time, NRC has been engaged in thedevelopment of a more explicit policy for nuclear
facility decommissioning. In September 1978, NRC
held a set of regional workshops to review with State
officials the specifics of the NRC plan. The plan was
modified in response to State comments, reissued, andsent to all the States for review. Followup workshops
to discuss (1) the technical studies completed following

Three Mile Island Waste Shipments
The Office of State Programs provides advancenotice of radioactive waste shipments from Three Mile
Island to all States through which the shipments arerouted (the low-level waste is bound for the Hanford,
Washington, low-level waste burial site).
Information on the shipments, such as nature ofwaste, carrier, radiation levels, etc., is conveyed immediately to the State liaison officer and the State
radiation control program director in each of theStates involved in the routing.

9International
Activities
NRC inspector checks containers destinedfor overseas shipment for possible escapeof radiation.

The NRC's international activities encompass for~mal exchanges of information and cooperation withother countries regarding radiological health and safety, administration of nuclear export and import licensing, efforts to deter nuclear proliferation, and theclosely related area of international nuclear
safeguards.
During fiscal year 1979, the NRC:
• Executed four new agreements with other countries and international organizations for the conduct of experiments or exchanges of informationin the nuclear safety research area.
• Maintained bilateral arrangements for regulatory
information exchange and cooperation with 17
countries and began or continued negotiations for
similar arrangements with seven other countries.
• Held policy and technical meetings with 571
visitors from 32 countries and four internationalorganizations. The increase of nearly 40 percent
in the number of foreign visitors during the year
was due mainly to concern over the accident at
the Three Mile Island nuclear plant in Pennsylvania.
• Issued 678 nuclear export licenses, of which 154
were major licenses, and received 709 new export
license applications.
• In its role of implementing the Nuclear NonProliferation Act of 1978, reviewed and provided
NRC views to the Executive Branch on 13 requests from four countries for approval ofretransfer of U. S. -origin spent nuclear fuel toother countries for reprocessing, and consulted
with Executive Branch agencies on several cases
involving export of technology associated with the
production of special nuclear material outside theUnited States.
• Continued to support domestic and international
efforts to develop and operate the nuclear fuel cy-

cle in ways that minimize the risks of nuclear profileration.
• Worked closely with the Executive Branch toassist the International Atomic Energy Agency instrengthening international safeguards.
• Issued revised proposed regulations to implement
the US/IAEA Safeguards Agreement when
ratified by the Senate. The Agreement provides
for the application to U.S. civil nuclear facilities
of IAEA safeguards.

Information Exchanges
Bllateral Arrangements
The NRC has entered into regulatory inform'ationexchanges and cooperation arrangements withregulatory bodies of 17 countries since 1974: Belgium,
Brazil, Denmark, France, the Federal Republic of
Germany, Greece, Iran, Israel, Italy, Japan, Korea,
the Netherlands, Spain, Sweden, Switzerland,
Taiwan, and the United Kingdom. While no new arrangements were concluded during fiscal year 1979,
negotiations were either begun or continued withregulatory bodies in Canada, Egypt, Finland, Mexico,
the Philippines, Yugoslavia, and the Union of Soviet
Socialist Republics.
The objectives of the arrangements are to:(1) Establish a formal channel of communication
with foreign regulatory organizations to assure
prompt and reciprocal notification of reactor
safety problems that could apply to both U.S.
and foreign nuclear facilities.
(2) Form a network for bilateral cooperation
related to public health and safety, safeguards,
and environmental protection.

184(3) Assist in developing an international consensus
on regulatory matters and safety standards andexperiments.

New Research Agreements(4) Provide assistance in improving nuclear healthand safety practices of countries importing U.S.
reactors.
Provisions of the arrangements typically call for thereciprocal exchange of regulatory information in theform of technical reports, correspondence, newsletters, meetings, training courses, and any other meansagreed upon. In some cases, they also provide for
future cooperation in reactor safety research and temporary assignments of personnel to agency headquarters and laboratory programs under the sponsorship of both parties.
Such arrangements are effective for 5 years, but
may be extended by mutual written consent. Five of
NRC's arrangements (those with Japan, France,
Spain, Sweden, and Switzerland) were either renewed
or were in the process of being renewed in 1979.

During 1979, the NRC executed four new
agreements involving nuclear safety research. A tripartite administrative agreement. was concluded between the Bundesminister Flir Forschung andTechnologie of the Federal Republic of Germany(FRG), the Joint Research Center of the Commissionof the European Communities, and the NRC to collaborate on a series of molten salt-water experiments.
This program will provide data for confirming
theoretical models used in the analysis of possible
steam explosions in light water reactors.
A 5-year agreement was concluded with the European Atomic Energy Community (EURATOM) covering a broad exchange of research information on bothlight water and fast reactor topics.
Two research agreements were concluded with theFederal Office of Energy of Switzerland (EAW). The
first agreement provides for the exchange of information and EA W's participation in the NRC's Loss-ofFluid Test (LOFT) program and reciprocal activities

In October 1979, a delegation from the Federal Republic of Germany (FRG), headed by Dr. Volker Hauff, Federal Minister for
Research and Technology (far right), visited the NRC to discuss
waste management and other nuclear safety topics with Commissioner John Ahearne (far left) and Dr. Joseph D. Lafleur, Deputy

Director, Office of International Programs. NRC maintains closeworkinp; relations with West German e:ovemment ae:encies,
through bilateral agreements for cooperation and research information exchange, as well as throusdt continuous contacts within
the International Atomic Energy Agency.

185

The accident at Three Mile Island in March 1979 resulted inmany requests to visit the site from foreign scientists and engineers.
The NRC staff arranged for such visits and accompanied thevisitors on several occasions. One such group is shown here about
to enter the Middletown, Pa., National Guard Anoory for abriefing on the TMI situation.

by NRC in the emergency core cooling thermalhydraulic experimental program and the loss-ofcoolant accident analysis program of the Swiss Federal
Institute of Reactor Research. The second agreement
provides for EAW participation in the NRC's Heavy
Section Steel Technology program in return for NRC
participation in the fracture mechanics research program in Switzerland. Each of the agreements has aterm of 4 years.
Additionally, the NRC renewed its participation in
the Halden Reactor Project for a period of 3 years. The
Halden project is sponsored by the Nuclear Energy
Agency of the Organization for Economic Cooperation
and Development (OECD) and is conducted by the
Norwegian Institut for Atomenergie. The program
covers fuel reliability and safety, fuel performancemodeling, and the use of process computers in plant
control.

protection. Some visits included tours of U.S. nuclear
facilities and national laboratories to observe NRC
safety activities and research programs.
An increase of nearly 40 percent in the number offoreign visitors over the total for the preceding year
was due to concern over the Three Mile Island nuclear
power plant accident. Most of the visitors were fromcountries with which NRC has bilateral, regulatory
and safety, and research arrangements. Numerous
delegations of reactor specialists were escorted to the
TMI site for discussions with NRC operating personnel.
A number of regulatory officials from other countries participated with the NRC staff in ongoing programs to gain experience in the U.S. regulatory process
and to contribute their expertise to various tasks overperiods of 1 month to 1 year. Participants in fiscal year
1979 were from Brazil, Israel, Korea, Mexico, thePhilippines, and Turkey.

Increase in Foreign Visitors
During fiscal year 1979, the NRC held policy andtechnical meetings with large delegations and individual visitors from foreign countries and organizations totaling 571 persons from 32 countries and fourinternational bodies. These included several 2- to
3-day discussions with foreign administrators of information and cooperation agreements with NRC as wellas with their designated representatives regarding
operational safety, safeguards, and environmental

IAEA Nuclear Safety Program
The NRC assisted the IAEA in augmenting its
nuclear safety program following the accident atThree Mile Island. The expanded IAEA program,which was approved by the IAEA Board of Governors
in June 1979, involves the addition of IAEA staff
specialists to work with developing countries on reactor safety matters, additional safety guidance to beprepared under the IAEA nuclear reactor safety standards program, special meetings to exchange informa-

186tion on significant safety incidents at power reactors,
and increased efforts on emergency plans andassistance. The NRC has provided, without cost to the
IAEA, the services of a senior staff member for one
year to assist in providing reactor safety advice to theregulatory authorities of developing countries with
nuclear power programs.
NRC and IAEA will co-sponsor a meeting early in1980 to discuss the actions being taken in the U.S. andother countries in response to the Three Mile Island accident. Preliminary briefings on the accident were
given by the NRC staff to international representatives
visiting the U.S. in April and May 1979.

Technical Assistance through IAEA
In coordination with the IAEA Technical Assistance
Program, the NRC continued to provide safety advice
and assistance to regulatory authorities of countries
embarking on nuclear power programs.
Several short-term reactor safety missions to support
the Brazilian National Nuclear Energy Committee,
begun in 1978 on behalf of the IAEA, were completed
in 1979. The IAEA has requested NRC's continued
assistance to Brazil in the areas of operator licensing
review, criticality preparations, power ascension, andhealth physics. Two short-term missions were carriedout by NRC staff experts to assist the National Nuclear
Safety and Safeguards Commission of Mexico regarding concrete and containment systems.
A safeguards training program for Korea, begun in
1978, was completed in 1979 and developed into amodel program for use by any country that may request NRC assistance in strengthening its state system
of material accounting and control.
Also, the NRC staff has made available two experts
for I-year assignments as advisors under IAEA
auspices to countries undertaking a strengthening of
their nuclear regulatory programs. One has been
assigned to the Philippines to advise on quality
assurance and structural engineering, and the other to
Mexico to advise on licenSing reviews.
Training Courses Held. NRC staff memberspresented lectures as part of three courses conducted in1979 on behalf of the IAEA by Argonne National
Laboratory's Center for Educational Affairs. These included:• A 4-week course in safety analysis review, held in
Seoul, Korea.
• A I-month course on nuclear power plant siting,at Argonne Laboratory.
• A 6-week interregional training course at
Argonne on safety and reliability in nuclear plant
operation, open to candidates from developing
countries.
The NRC also provided lectures for an IAEA course
at the Karlsruhe Research Center in the Federal
Republic of Germany.

Cooperation with the OECD
NRC is represented on several committees of theOECD's Nuclear Energy Agency, with major effort
ceiltered in the Committee on the Safety of Nuclear Installations (CSNI) and its Licensing Subcommittee.
NRC helped plan and carry out the CSNI response to
the accident at Three Mile Island, including efforts
toward establishing an effective system for reporting
information concerning safety-significant reactor incidents occurring in any of the 24 member countries.
NRC also played a prominent role in an October 1979
Specialists Meeting of Regulatory Review in the Licensing Process, held in Madrid, Spain, and co-sponsoredby the CSNI Licensing Subcommittee and the Spanish
Junta de Energia Nuclear.
NRC senior staff also participated in activities of the
NEA standing committees on Radiation Protection
and Public Health and on Waste Management, and of
the NEA Ad Hoc Group for the Study of Administrative and Financial Aspects of Long-Term Managementof Radioactive Waste.

Export/Import Matters
And Nonproliferation
The NRC continues to perform a vital role indeveloping, supporting, assessing, and implementing
U.S. nonproliferation policy. The Nuclear Nonproliferation Act of 1978 (NNPA) provides a policy
framework for the discharge of responsibility by the
NRC and the Executive Branch in (1) ensuring that
nuclear export activities are conducted promptly andin conformance with national security and specificcriteria set forth in the Act, (2) strengthening IAEA
safeguards, (3) improving physical protection
measures, (4) improving nuclear fuel assurances toother countries, (5) renegotiating Agreements for
Cooperation, (6) evaluating alternative fuel cycles,and (7) formulating spent fuel disposition policy.
In addition to NRC's direct export licensing activities, the NNP A requires Executive Branch agencies
to consult formally with NRC on nuclear exportrelated activities under their purview, including:• Negotiation of new and revised agreements forcooperation (State Department and Departmentof Energy).
• Nuclear technology exports (DOE).
• Foreign distribution of nuclear material (DOE).
• Negotiation of contracts for the supply of nuclear
materials and equipment (including enrichmentservices to foreign recipients) (DOE).
• Consideration of requests to retransfer
U.S.-supplied nuclear material and facilities(DOE).

187• Consideration of requests to reprocess irradiated
U.S.-supplied nuclear fuel (DOE).
• Other "subsequent arrangements" as defined in
section 131 of the Atomic Energy Act of 1954, as
amended.
• Exports of nuclear-related commodities by the
Department of Commerce.
The NRC's nonproliferation role became increasingly visible during the year, both in overall activities
in the international sphere and as a member of theinteragency Subgroup on Nuclear Export Coordination (SNEC) which considers many significant or controversial nuclear export matters to facilitate appropriate actions. (NRC participates in SNEC only inan observer capacity when the Executive Branch is formulating its position on individual export license applications filed with NRC.)

Retransfers for Reprocessing
In its role as a member of SNEC, the NRC in fiscal
year 1979 provided its views on 13 requests toretransfer U.S.-supplied nuclear material to othercountries for reprocessing: four from Spain, five fromJapan, two from Switzerland, and two from Sweden.
While the NRC generally has not opposed such requests, the Commission has stressed strict adherence to
NNPA requirements and the Presidential criteria promulgated in the TEPCO and Kansai cases (see 1978
NRC Annual Report, p. 150). It was stated in thesecases that such requests could be approved based on
the existence of a reprocessing contract that antedates
April 1977, or based on the physical need to remove
the spent fuel from congested storage pools.
In its review of such requests, the Commission has
expressed its concerns to DOE and has sought to encourage development of the overall U.S. policy regarding deferral of reprocessing activities. Upon completion of the International Nuclear Fuel Cycle
Evaluation (see discussion later in this chapter) in early 1980, the NRC anticipates active participation inestablishing new policy and approval guidelines in thisregard.

Technology Transfers
The NNPA requires the Executive Branch agencies
to consult with NRC regarding exports of technology
involving directly or indirectly the production of
special nuclear material outside the United States,
which require specific authorization by the Secretaryof Energy and are controlled by DOE under its regulations in 10 CFR Part 810. Under Part 810, all exportsof such technology to communist countries or exports
related to reprocessing, enrichment, heavy water production, and plutonium fuel fabrication to any
destination require specific authorization unless theinformation is publicly available in published form.

During 1979, the NRC considered several suchcases, including the transfer of enrichment technology
from West Germany to Brazil and the sale of CANDU
reactor components to Romania. NRC is encouraging
the Executive Branch to develop an overall policy on
the export of laser isotope separation technology.

Agreements for Cooperation
The renegotiation of agreements for nuclear
cooperation, called for in the NNPA, continued in
1979. A principal feature of the renegotiation process
has been to make reciprocal for both the U.S. andsome of its trading partners the provisions regarding
physical security and the storage, retransfer, andreprocesssing of spent fuel.
Under the lead of the Department of State, and in
consultation with other U.S. agencies, including the
NRC, renegotiated agreements have been concluded
with Australia and the IAEA. Agreements with Indonesia, Sweden, Norway, Finland, Peru, Morocco,
Columbia, Korea, Japan, and Egypt, were in various
stages of completion at the end of the fiscal year.

Nuclear Fuel Cycle Evaluations
The NRC continued to participate in support of activities associated with both international anddomestic evaluations of nuclear fuel cycle systems
aimed at reducing proliferation risks. These programsare:• The International Nuclear Fuel Cycle Evaluation(INFCE), an effort begun in October 1977 anddue to end in March 1980 in which 53 countries
and four international organizations areevaluating means of developing and operating thenuclear fuel cycle in a manner to minimize therisks of proliferation.
• The Nonproliferation Alternative Systems Assessment Program (NASAP), a related U.S. effort being conducted by DOE which is providing
technical data and input to INFCE.
The 1979 NRC Authorization Act (P.L. 96-601)
directed the agency to monitor and assist, as requested, both the INFCE and NASAP studies and toreport to Congress on their status semiannually
through calendar year 1980, and annually thereafter
through 1982. The first suc,h report, covering activities
through June 30, 1979, was sent to Congress in
December 1979. It noted that, because the NASAP and
INFCE reports had not yet been released and becauseof the strain on agency resources caused by the Three
Mile Island nuclear plant accident, NRC was able toprovide only limited analysis of the status of theevaluations in the first report. With Congressional approval, NRC had delayed much of a carryover of

188==================================================$800,000 in fiscal year 1978 funds earmarked for alternative reactor and fuel cycle research in order to meetresource needs precipitated by the Three Mile Island
accident.
The NRC has reviewed and provided comments on
preliminary NASAP documents furnished by DOE on
the six principal reactor concepts being considered:light water, heavy water, light water breeder, hightemperature gas-cooled, gas-cooled fast breeder, andliquid metal fast breeder reactors. Other DOE
documents reviewed included those discussing theassociated fuel cycle facilities and safeguards considerations for the alternative cycles.
NRC is sponsoring technical work by outside consultants to better understand licensing issues unique to
the NASAP reactor conceptual designs and to obtain
information useful for establishing future programs
and policies regarding safeguards for fuel cycles
developed commercially. In addition to technicalwork supporting the alternative reactor and fuel cycle
effort, NRC maintains a significant research programin fast breeder reactors and a limited research effort inadvanced converter reactors.
The main proliferation concerns center on the backend of the fuel cycle-reprocessing and fabrication ofrecycle fuel, and waste management. All fuel cycles
considered, except once-through, require reprocessing
and recycling of fissile materials, and more information is needed for appropriate NRC assessment of therelative proliferation risks.
Additional information and analysis also are required to reach judgments on the relative safety,
safeguardability, environmental impact, and Hcensability of the alternative reactor and fuel cycle concepts that have been reviewed.

Pacific Basin Fuel Storage
The concept of storing spent nuclear fuel in centralized regional facilities as a means of enhancing
energy security of countries dependent upon nuclear
power and advancing shared nonproliferation objectives is being explored by the United States. Specifically, the Department of State and DOE are continuing
to pursue with the government of Japan the issue of a
joint feasibility study on an interim spent fuel storage
facility in the Pacific Basin.
Observers from the NRC staff attended several
interagency planning meetings during the year andcontributed to the consideration of technical andregulatory issues raised by other agencies regarding
the possibilities of Pacific Basin fuel storage. Discussions with interested Pacific Basin countries are continuing. NRC plans to offer informal staff comments
and assistance to the Executive Branch agencies engaged in the study.

NRC Views on Nonproliferation Role
Section 602 of the Nuclear Nonproliferation Act requires the Commission and DOE to include in theirannual reports to the Congress "views and recommendations regarding the policies and actions of theUnited States to prevent proliferation which are thestatutory responsibility of those agencies ... "Progress has been made in expediting the processingof export license applications through NRC's revised
export/import regulations (10 CFR Part 110, issued in
May 1978) which incorporate licensing criteria and requirements of the nonproliferation legislation.
However, delays are unavoidably entailed in thereview of significant export cases by at least six
separate U.S. Government agencies, particularly
when complex issues and differing interpretations of
the applicable export review criteria are involved.
Since enactment of the NNPA, the Commission has:
(1) Delegated to the staff authority to processroutine low-enriched uranium fuel export
license applications without referral to the Executive Branch.
(2) Established a limited policy to license several
nuclear fuel reload exports at one time, thereby
reducing the number of applications that need
to be filed and processed.
(3) Adopted the approach, authorized by sections
126a(2) of the Atomic Energy Act of 1954, as
amended, of issuing export licenses upon a finding of "no material changed circumstances"from those existing at the time the last license
application for export to the same country wasapproved using NNPA procedures, thus
eliminating repetitive analyses of some applications.
(4) Approved filing of consolidated export license
applications covering more than one shipmentof similar material or equipment to various
countries and consignees.
(5) Streamlined internal Commission and staff
reviews of exports handled both by the NRC andother agencies.
In November 1978, the NRC proposed a standard
format for the Executive Branch's analyses of nuclear
export license applications in an effort to reach agreement on those matters which the Commission believes
should be addressed by the Executive Branch. Discussions were nearing completion at the end of 1979.
The Commission continues to believe further
clarification is needed of NRC's consultation role on
nuclear export matters under the purview of the Executive Branch-particularly with respect toretransfer requests involving reprocessing, on which
the Commission has provided several comments to the
Executive Branch.

189

The issues of the adequacy of IAEA safeguards applied to nuclear exports and NRC needs for moredetailed information concerning safeguards implementation abroad continue to concern the Commission. During the year, the NRC worked closely with
the Executive Branch in the continuing effort to improve international safeguards. (See discussion
below.)EXPORT LICENSING ACTIONS
During the fiscal. year ending September 30, 1979,
the NRC issued 678 export licenses and amendments toexisting licenses, and received 709 new export license
applications, including requests for amendments. Of
the 678 licenses issued, 154 were major licenses which
are listed in the accompanying table in three
categories: special nuclear material, source material,
and reactors. The 524 export licenses considered to beminor inclucled 162 for small quantities of special
nuclear material, 39 for source material, 76 for
byproduct material, and 247 for components. (NRC
also issued 53 import licenses, including amendments,
and received 43 new import license applications.)
Sixteen different nations received U.S. shipments ofspecial nuclear material under major export licenses
during the year. In addition, seven nations received
major quantities of source material, and three nations
received a reactor facility. No licenses were issued during the period for the export of large quantities of
plutonium.
Two significant export license cases are discussed
below.

Philippines Reactor Project
A reactor export application dating from 1976 andsubsequent license applications for the export ofassociated components and fuel for a nuclear power
plant at Napot Point in the Philippines drew substan. tial public attention-mainly over safety-during
1979.
An application by Westinghouse Electric Corporation (XR-120) for a license to export a nuclear facility
to the Philippines was filed in November 1976. In
August 1978, Westinghouse filed another application
for the export of several reactor components(XCOM-0013) needed to allow completion of the project, on which construction had been started. This action, which formerly would have been handled by the
Department of Commerce, was transferred to the
NRC under provisions of Section 309 of the Nuclear
Nonproliferation Act of 1978 which transferredauthority for licensing exports of nuclear components
from Commerce to the NRC. In February 1979,
Westinghouse also filed a license application(XSNM-1471) with NRC for the export of 121,000
kilograms of low-enriched uranium for the initial core
and three reloads for the Philippines facility. The Executive Branch has completed its review of these casesand has recommended that all be approved.
In June 1979, President Marcos of the Philippines,
citing concerns that the plant site may be too close toan active earthquake zone and to possibly volcanic
areas and also that the Westinghouse design may be
faulty, suspended construction of the nuclear plant
until issues concerning the safety of the operation
could be resolved. Subsequently, the Philippines
government established a Commission (the "Puno
Commission") to investigate these concerns. In
November, the Puno Commission released its findings,
which essentially concluded that the plant site was acceptable, but that serious concerns still remained
regarding the safety of Westinghouse's design. Basedupon these findings, President Marcos continued inforce the suspension of construction activities.
On April 19, 1979, the Center for Development
Policy, Jesus Nicanor P. Perlas III, and the Philippine
Movement for Environmental Protection filed a petition with the Commission requesting a public hearing
on the proposed export licenses. Petitioners requested
that the hearing focus on issues related to environmental, health and safety impacts that the proposed reactor would have upon the Philippines, including thepotential effects on U.S. citizens residing there. On
October 19, 1979, the Commission issued an order inviting petitioners and other members of the public tosubmit written comments to the Commission addressing the issues of:(1) Whether the Commission has jurisdiction to examine health, safety, and environmental impacts in a foreign country arising from the construction and operation of an exported nuclear
reactor.
(2) Whether its health, safety or environmental
review of export license applications is limited
to the consideration of those issues with U.S .
common defense and security implications, orwhether the legal principles would permit or require the Commission to examine such mattersas part of its licensing review.
(3)" What issues arising from the application to export a nuclear reactor to the Philippines should
the Commission examine in any future public
proceeding.
(4) What procedural format should be adopted forconsidering such issues if they are found to lie
within NRC jurisdiction.
(5) If health, safety and environmental aspects of aU.S.-supplied facility are to be evaluated in the
NRC export licensing process, in what manner190

Table 1: Major Nuclear Export Licenses(Major Licensing Actions Taken by NRC-October 1, 1978 through September 30, 1979)U

SPECIAL NUCLEAR MATERIAL (One or more "effective kilograms as defined in 10 CFR 70.4(t))
LicenseeTransnuclear
Transnuc1ear
Transnuc1ear
Marubeni America
Mitsui & Company
Mitsui & Company
Marubeni America
Transnuc1ear (rountine reload)Westinghouse
Westinghouse

Kilograms
oj Uranium
15.976

Transnuclear
Transnuclear

22.556.21.053
Increase maximum enrichment
Increase maximum enrichment
Increase maximum enrichment
Increase maximum enrichment
14,151
66,752
Extend expiration date;change licensee's address;
Additional
65,939243,000
1.07 gm
Extend expiration date
38.67544.40020 gmsExtend expiration date
Extend expiration date;delete conditions
Extend expiration date
Additional
.245
.930
35.423
i09,406.i70
107,386
Additional
36,443.1
Additional
50,037
198
779
36,78337,160
9,955
10,27026,222
9,679
11,191
Add Intermediate Consignees;delete another
Increase maximum enrichment
Add party to export;change licensee's name11,584.75
Extend expiration date

Transnuclear
Westinghouse
Transnuclear
Transnuclear
Transnuc1ear
Transnuc1ear
Transnuc1ear
Transnuc1ear
Transnuc1ear

Extend expiration date
Increase maximum enrichment
11,05634,80149,656
37,16024,062
22,295.905
5.297

General Electric
GETSCO
General Atomic
GETSCO
General Electric
General Electric
General Atomic
Transnuc1ear
Westinghouse (routine reload)
Transnuc1ear
Exxon Nuclear
Exxon Nuclear
Transnuc1ear
Transnuclear
Mitsubishi
General Electric
Mitsubishi
Edlow International
Mitsubishi
Mitsubishi
Mitsubishi
Transnuc1ear
General Electric
General Electric

E nrlchment %93 . 3
93.3
93.3

3.35
3.49

4.0
93.5
93
20
93

93.3
20.0
93.30
3.30
3.25

Country oj
Destination
Netherlands &W. Germany
Netherlands
Netherlands
Japan
Japan
Japan
Japan
Netherlands
Yugoslavia
Rep. of Korea

Date Issued

10/13/78
10/13/78
10/13/78
10/31178

10/31178
10/31178
10/31178

11/02/78
11/09/78
11/17/78

Switzerland

11/17/78

Japan
Romania

11/22178
11/22/78

Switzerland
Spain

11/22/78
11/22/78

Italy
Romania

11/27/78
11/29/78

Japan
SwedenW. Germany
Sweden12/01178
12/07/78
12/26/78
12/26/78

W. Germany

01102/79

CanadaCanada
Japan
Japan
Japan
Japan
Japan
J-apanJapan
Netherlands

01104/79
02/01179
02/02i79.
02/02/79
02/02/79
02/02179
02/02/70
02/02/79
02/02/79
02/09/79

Spain·
Japan

02/16/79
02/16/79

Switzerland
W. Germany
Netherlands
Netherlands
Spain
FranCe·W. Germany
FranceW. Germany
W. Germany
W. Germany
Canada02/16/79
02/16/79

2.90
3.20
93.3
4.05
3.25
3.65
2.65
3.15
2.85
3.15
2.85

3.55

4.335
3.25
3.30
3.0
3.4
3.0
93.3

02116/79
02/22/79
02/23/79
02/23/79
02/23/79
02/23/79
02/23/79
02/23/79
02/26/79

191

SPECIAL NUCLEAR MATERIAL (One or more "effective kilograms" as defined in 10 CFR 70.4(t))
LicenseeTransnuclear
General ElectricTransnuclear
Transnuc1ear
Transnuclear
Transnuclear
General Electric
Edlow International
Union Carbide
Westinghouse
Westinghouse
General ElectricTransnuclear
Westinghouse

Transnuclear
Transnuclear
Transnudear
Transnuclear
Mitsui & Company
Transnuc1ear
General ElectricTransnuclear
Exxon NucIear Co.
Transnudear
Transnuclear
TransnucIear
Edlow International
TransnucIear
Mitsubishi
Exxon NucIear
TransnucIear
Transnuclear
TransnucIear
Transnudear
General Electric
General Electric
General ElectricTransnudear
Westinghouse
Transnuclear
GETSCO
General ElectricTransnuclear
Mftsubishi
General ElectricTransnucIear
General Ek'Ctric
Mitsui & Company
Mitsui & Company
Edlow International
Transnudear
Mitsui & Company

Kilogramsof Uranium

Country of
Destination

Date Issued

Belgium
MexicoW. Germany

02/26179
03/02179
03/14179

W. Germany

03/14179

W. Germany

03/14179

japan
Italy

03/14179

2.71
93.16

India
FranceBrazilFranceMexico03/23179
03/26179
03/26179
03/22179
04/03179

93.3

FranceSwitzerland

04/09179
04/09179

Netherland__
Netherlands
SwedenSwedenjapan
W. Germany
FranceW. Germany
Belgium
W. Germany
Belgium
CanadaSwedenjapan

04/11179
04/11179
04/11179
04/11179
04/13/79

Enrichment %10,794
377,600
Delete Other Party to Export;
add Shipper of Record
Delete Other Party to Export;
add Shipper of Record
Delete Other Party to Export;
add Shipper of R<..'Cord
4.900
Extend expiration date;delete condition
16,803.6
7.33
Extend expiration date
2.48 Pu; 61 natural U
Change name of 2nd Ultimate
Consignee
23.058Remove Intermediate Consignee;
extend expiration date;delete conditions;
Additional
23,34720.05019.8
22.05517.043
1,897
22.0
Add condition
12,71455,40018,951
10,581
4.23
61,750
Delete Other Party to Export;
submit another for transport only
16,341
Extend expiration date
22,090.900
10,225Extend expiration date
Extend expiration date
Extend expiration date
239,000
Delete Ultimate Consignee
14,101
Extend expiration date;delete condition;change maximum enrichment
12,283.11
Extend expiration date
Add conditions
Add conditions
11,983
11,965
11,585.0
5,875
26,902
Include enrich<..>d uranium as U0 2
13,12522,090.0
157,382

3.40
4.0

93.3

3.55
93.3
93.3
93.3
93.3
3.85
93.3
3.25
3.5
3.35
3.4
93.3
3.55
3.33
3.35
3.0

4.0
3.25

4.3

'2.85
3.1
3.55
3.1
3.95
4.52
3.35
2.55

03/19179

04117179
04/26179
04/27179
04/27179
04/27179
04/2717905/08/79
05/09179
05/09179

japan
W. Germany
Switzerland
W. Germany
W. Germany
japan
japan
TaiwanTaiwanSwedenSpain

05/10179
05/24179
05/29179
06/05179
06/05179
06105179
06106179
06/08179
06113/79

Francejapan
France
japan
japan
japan
Switzerland
japan
japan
japan
Italy
Switzerlandjapan

06/25179
06/27179
06/28/79
06/28179
07/03179
07/03/79
07/03/79
07113/79

06/20/79
06/25179

07113179
07118179
07/20179
07/27/79
08/02179

192

Table 1: Major Nuclear Export Licenses-Continued(Major Licensing Actions Taken by NRC-October 1, 1978 through September 30, 1979)
SPECIAL NUCLEAR MATERIAL (One or more "effective kilograms" as defined in 10 CFR 70.4(t))
LicenseeTransnuclear
General Electric
Marubeni America
General Electric
Marubeni America
Marubeni America
Transnuclear
Transnuclear
Edlow International
Edlow International
Edlow International
Mitsui & Company
Edlow International
Edlow International
Edlow International
Westinghouse
Westinghouse
Mitsui & Company
Exxon Nuclear

Kilograms
oj Uranium

Enrichment %10,79429,52630,382
18,452
25,723
3,799
17,5893.810
Extend expiration date
Extend expiration date
Extend expiration date
3,808
Extend expiration date;delete conditions
Extend expiration date;delete conditions
44,870Extend expiration date
122,220
9,214
21,237
Extend expiration date;change licensee's address

3.4

3.65
2.97
3.65
3.90
3.90
3.35
93.3

3.95

3.55
3.55
3.95
3,95

Country oj
Destination

Date Issued

Belgium
japan
japan
japan
japan
japan
Netherlands
Austria
SwedenSwedenSwedenjapan
Sweden08/08179
08/13179
08/13179
08/13179
08/14179
08/14179
08/23179
08/27179
09/05179
09/05179
09/05179
09/06179
09/06179

Sweden09/06179

SwedenSwedenSwitzerlandjapan

09/06179
09/11179
09/11179
09/20179

Sweden

09/21179

SOURCE MATERIAL
Material

LicenseeN L Industries
8,164.746
RMI Company
Transnuclear
N L Industries
Nuclear Metals, Inc.
Edlow International
Mitsubishi
Army Marl. & Mech.
RMI Company

304,818
9,072
23,000538,511250
90,039

Aerojet55,344
Rhone-Poulenc
Edlow International
NUS Corporation
NUS Corporation
Edlow International
Edlow International
Edlow International
Transnuclear
U.K. Treasury &Supply Delegation
Transnuclear

100,335
10,165
10,00016,8001,000,000
129,243
125,500
500
1,200
126,000

Additional
kgs. depleted uranium
Extend expiration datekgs. depleted uranium
Additional
kgs. depleted uranium
kgs. depleted uranium
Extend expiration datekgs. natural uranium
kgs. depleted uranium
Additional
kgs. depleted uranium
Additional
kgs. depleted uranium;
extend expiration datekgs. thorium
kgs. natural uranium
kgs. depleted uranium
kgs. depleted uranium
Delete Ultimate Consignee
Delete Ultimate Consignl.'C
kgs. natural uranium
kgs. natural uranium
contained in 152,410 kgs.
of U3 0 S
kgs. depleted uranium
kgs. depleted uranium
Increase material exported to
kgs. depleted uranium
Increase quantity to
kgs. depleted uranium

Country oj
Destination

Date Issued

Italy

10/13178

CanadaSwitzerland

10/16178
11103178

CanadaUnited Kingdom
CanadaCanadaUnited Kingdom
Canada01103179
03/01179
03/21179
04/09179
05/09179
05/09179

Canada05/22179

France05/24179

Rep. of China
Rep. of China
TaiwanTaiwanUnited Kingdom
West Germany

06/08179
06/08179
06/13179
06/13179
06/21179
07/27/79

France07/27179

United Kingdom

09/25179

France

09/06179

193

REACTORS
LicenseeWestinghouse
Westinghouse
GETSCOGETSCO

Facility Description

Two 2,785 MWT PWR KOR1-3 and
KORI-4; value of items $200,000,000
Extend expiration date; changelicensee's address
Kaiseraugst Nuclear Power
Station/BWR/2894 MWT'; value$23,000,000
Add Additional Authorized' Exporter;

Country oj
Destination

Date Issued

S. Korea

10/04/78

S. Korea

10/16/78

Switzerland

11/20/78

Japan

11/27/78

Spain
Spain

01103/79
01123179

MexicoMexicoSweden02/01179
02/01179
02/02/79

Brazil02/02179Japan

03/26/79

RomaniaJapan

05/10179
05/14179 .

Rep. of China

06/08/79

TaiwanSpain

06/13/79
06/27/79

BrazilTaiwan08/23/79
09/11/79

del~t(:l.condition

Westinghouse
GETSCO
General Electric
General Electric
Westinghouse
Westinghouse
General Electric
General Atomic
General Electric
Westinghouse
Westinghouse
General Electric
Westinghouse
General Electric

Add Intermediate Consignee
Extend expiration date;change licensee's address
Add another party to export
Add another party to export
Extend expiration date: increasevalue to $28,000,000
Conform early license with recent
licenses
Change addresses of Intermediate
Consignees; add 2 other parties toexport; delete Intermediate Consignee
Extend expiration date
Change addresses of Ultimate andIntermediate Consignees
2785 MWt PWR
Taiwan Power Co. Units 5 & 6
Delete Ultimate Consignee
Change address of Ultimate
Consignee
Extend expiration date
Export instruments for 2 BWRs: add
other parties to export; change otheraddresses

should the review be conducted differently from'domestic reactor licensing proceedings.
(6) Whether there are any factual or legal considerations which would justify different NRC
health, safety or environmental reviews forsome export license applications than for others.
After reviewing the submissions received, the Commission is expected to issue a second order which may
or may not result in additional proceedings on the
merits of the proposed Philippine exports. (See
Chapter 1 for further Commission action in January
1980.)In a separate development in this case, in an effort
to force issuance of the license for the, export of thenuclear components, Westinghouse filed a suit against
the NRC in August 1979. In the suit, Westinghouseargued that NRC's determination not to act on thecomponents license until questions regarding the safety of the site are resolved constituted action unlawfully
withheld and unreasonably delayed. On August 30,
the U.S. District Court of the District of Columbia
ruled in favor of the NRC.

Tarapur (India) Proceeding
The NRC Annual Reports for 1976, 1977, and 1978
contain detailed discussions of the history of the Commission's deliberations on the license applications toexport low-enriched uranium to India for use in the
Tarapur Atomic Power Station. In March 1979, by athree to two vote, the Commission authorized issuanceof license XSNM-1222 to India, covering 16,804
kilograms of low-enriched uranium. The Commission
majority concluded that this license application met
all the requirements for issuance under the Atomic
Energy A9t of 1954. The dissenting Commissioners
were of the view that, because of unique features in
the United States-India Agreement for Cooperation,
India's failure to adopt full. scope safeguards, and thelack of additional assurances covering the eventual
fate of U.S.-supplied fuel, a finding was precluded
that the United States had adequate assurances thatmaterial supplied to India would be kept under IAEA
safeguards, not be used to develop nuclear explosive
devices, and not be reprocessed or transferred toanother nation without prior U.S. approval.

194

In September 1978, Edlow International filed an
application for the export of another 19,858.8
kilograms of low-enriched uranium to India for conversion and fabrication into fuel elements for the
Tarapur Station. Favorable Executive Branch views
were provided in July 1979. However, in view ofsubsequent developments in India the Commission has
deferred action on this request until the political situation in India has stabilized sufficiently to permit an
updated Executive Branch judgment concerning India's nuclear policies and intentions. An additional application (XSNM-1569) for fuel export to India was filed in August 1979. This is still under review by the Executive Branch.

ENVIRONMENTAL EFFECTS
OF EXPORTS
Unified Interagency Procedures
On January 4, 1979, the President issued Executive
Order 12114, "Environmental Effects Abroad of Major Federal Actions," effective September 4, 1979,·
which directed all Federal agencies to join in developing unified implementing procedures for environmental reviews of certain nuclear exports.
The Order applies to several types of major Federal
actions, including: (1) actions significantly affecting
the environment of the global commons (e. g., theoceans and Antarctica), (2) actions significantly affecting the environment of a foreign nation not participating in the action, (3) actions affecting the environment of a foreign nation which provide to thatnation a product or physical project producing toxic
emissions which are strictly regulated in the U.S. (e.g.,radioactive substances), and (4) actions whichsignificantly affect natural or ecological resources ofglobal importance. In the nuclear area, the Orderspecifically applied to Executive Branch actions providing to a foreign nation a nuclear production,utilization, or waste management facility. Exports ofnuclear fuel, however, are excluded from the requirements of the Order.
The Order calls for the preparation of certain typesof documents to aid in assessing the significance of theenvironmental consequences stemming from the applicable Federal actions. A draft of proposed interagency procedures was distributed at the end of
August to all Federal agencies, including the NRC, forformal review and concurrence or comment. Following approval by Executive Branch agencies, the procedures were issued on November 5, 1979 and published in the Federal Register on November 13. The
NRC has not formally commented on the procedures.
The first export case for which the State Department has prepared an environmental document consistent with Executive Order 12114 is the reactor export
to the Philippines.

Research and Development Reactor Exports
The NRC assisted the State Department in preparing a document which examined the environmental
aspects fo nuclear research and development facility
exports. Its purpose was to determine, in light of theongoing and prospective cooperation in research anddevelopment activities, whether the nature andlor extent of environmental impacts on the global commonsor the U.S. from such activities require the prepara.,tion of environmental documents or impact statements
pursuant to the National Environmental Policy Act of1969. The analysis drew on evaluations alreadypresented in other environmental assessments, particularly the Final Environmental Statement of U. S.
Nuclear Power Export Activities (ERDA-1542). The
analysis concluded that neither the operation of, norpotential accidents associated with, exported researchand test reactors would have a significant impact on
the environment of the global commons or the U. S.
and that none of the impacts is likely to exceed those
associated with the power reactors considered in
ERDA-1542.
The report also examined the possible environmental impacts of the program for the return, temporarystorage, and reprocessing of foreign spent research andtest reactor fuel, and concluded similarly that no
significant impacts are likely to result from this program.

International Safeguards
International safeguards continued to draw
substantial attention of the NRC in fiscal year 1979. In
addition to responsibilities associated with the licensing of exports of nuclear materials and facilities, which
require NRC to consider the implementation of international safeguards in recipient countries, NRC wasinvolved during 1979 with the voluntary applicationof international safeguards at civil nuclear facilities in
the U.S.

US/IAEA Safeguards Agreement
The NRC devoted further attention in 1979 toactivities. related to the USIIAEA Safeguards Agreement, the voluntary U.S. offer to permit application ofinternational safeguards by the International Atomic
Energy Agency to civil nuclear facilities in the U.S.
Under this agreement, the U.S. would provide to the
IAEA safeguards information about U.S. civil nuclear
facilities "not of direct national security significance."
From these, the IAEA would select a number offacilities for the implementation of full safeguai:as inspections by IAEA inspectors. Implementation of theagreement will fulfill a 1967 Presidential offer to apply

195

IAEA safeguards to V.S. civil nuclear facilities inorder to demonstrate to other nations-particularly
the developed nonnuclear weapon states-that the application of international safeguards would not result
in commercial disadvantages. The Vnited Kingdom
and France, both nuclear weapon states, have madesimilar voluntary offers.
The two major developments during 1979 were the
Senate hearings on the proposed Safeguards Agreement and the republication for public comment of the
proposed NRC regulations to implement the treaty
with respect to NRC licensees. The Subcommittee on
Arms Control and International Operations of the
Senate Committee on Foreign Relations held hearingson the VS/IAEA Safeguards Agreement on June 22.
Officials from the Department of State, Arms Control
and Disarmament Agency, Department of Energy,
and NRC testified on the intent, scope, and likely impacts of the treaty. Representatives of the V.S. nuclear
industry joined the government officials and testified
in favor of the treaty. Additional hearings were heldby this Subcommittee in December 1979.
Proposed Part 75 of NRC's regulations to implement
the Safeguards Agreement, initially published for
public comment in 1978, was republished in revised
form in July 1979 to afford further opportunity forlicensee and public participation. In preparing therevised proposed regulations, NRC endeavored to state'
the requirements as clearly as possible and to provide
for licensees to take an active role, especially in the
preparation of "Facility Attachments" (which define
the safeguards to be applied at specific facilities),while at the same time making sure that the objectivesof the Agreement are implemented effectively.
In 1979, the NRC established an internal Safeguards
Agreement Implementation Group to coordinate allagency activities directed at implementing the Agreement. These activities have included an assessment of
the costs that the Agreement will impose on licensees,
revision of NRC/DOE nuclear material reporting
forms and procedures to make them consistent withthe requirements of the Agreement, and developmental work to facilitate preparation of Facility Attachments as soon as the Agreement comes into force.

Export Licensing Information Needs
As discussed in the 1978 Annual Report (pp.
153-154), the NRC safeguards staff has identified aneed for additional information on' the implementation of international safeguards for use in reviewing
export license applications. During fiscal year 1979,
the NRC and the Department of State began developing an approach to meet the needs of the Commission
which will be consistent with overall V.S. policy oninternational safeguards and the nonproliferation of
nuclear weapons.

The IAEA Safeguards Implementation Report (SIR)
for 1978, received by NRC in 1979, identified thesafeguards implementation problems that existed during calendar year 1978 and the corrective activities
undertaken by the IAEA. A number of the problems
were unchanged from those identified in the reports
for 1976 and 1977. Reasons for the persistence of these
problems include the resource constraints which the
IAEA faces. These constraints include both a lack ofinspection personnel and equipment, and the difficulties of expanding safeguards implementation
apace with the rapidly growing number of nuclear
facilities subject to safeguards.

Support of International Safeguards
During fiscal year 1979, NRC continued to work
closely with the Executive Branch on a number of activities designed to assist the IAEA in strengthening
international safeguards, including:• Participation in DOE's Program for Technical
Assistance to IAEA Safeguards. NRC's major contributions consisted of technical reviews of the activities and the provision of experts without costto the IAEA.
• Providing technical assistance to a foreign country in the development of its national system ofmaterial accounting and control, and the offer ofsimilar assistance to other countries on request.
• Working with the IAEA and Executive Branchagencies to provide a training course in the V.S.
for foreign officials who are responsible forestablishing and managing their countries'
national systems of material control and accounting.
• Participation in the V. S. Interagency Action Plan
Working Group to strengthen IAEA safeguards.
• Provision of direct technical assistance to the
IAEA.

Other Activities
Other activities related to the areas of internationalsafeguards and physical security of nuclear materials
which NRC undertook during the year included:(1) Participation with other V.S. agencies in
evaluating the safeguards aspects of technicalpapers prepared by International Nuclear Fuel
Cycle Evaluation (INFCE) working groups.
The NRC also provided experts to the INFCE"Safeguards Crosscut Group" in support of thiseffort.
(2) ParUcipation in meetings, both in the V.S. andabroad, with foreign experts on internationalsafeguards and physical security matters to exchange views and information.

196(3) Assignment on a long-term basis of NRC
safeguards technical experts to the lAEA staff in
Vienna.
(4) Participation in the U.S. negotiating team on
the Convention on the Physical Protection of
Nuclear Materials. This treaty, a U.S. initiative,
on which negotiations were concluded in Oc-

tober 1979, establishes the agreement of the international community on the appropriate
levels of physical protection to be accorded tonuclear materials during internationaltransport. The treaty also is designed tofacilitate international cooperation in thephysical protection of nuclear material.

10Standards
Development
NRC standards include specification andtesting criteria for measurement devices.

NRC standards provide for protection of the public
and nuclear industry workers from radiation, thesafeguarding of nuclear materials and facilities fromtheft and sabotage, and protection of the quality of theenvironment in nuclear activities. Thus, the development of standards cuts across the range of the NRC's
activities and requires close interaction between the
Office of Standards Development and the agency's
other program offices.
While many of the standards issued or worked on
during fiscal year 1979 are discussed in this chapter,some are discussed elsewhere in this Annual Reportunder the topics to which they relate (e.g., transportation in Chapter 4 and safeguards in Chapter 5).

CONCERNS OF HIGH PRIORITY
Current issues of high priority in standards development include:
Standards Development After TMI. Lessons learned
as a result of the accident at Three Mile Island (TMI)
will have a substantial impact on the NRC standards
program for fiscal year 1979 and succeeding years.
Major efforts to incorporate into NRC standards a consistent treatment of fission product release caused byfuel failure, and on improving NRC standards for
emergency planning and for nuclear power plant
operations, are underway. In response to recommendations from the NRC staff and from other groups thathave been investigating the TMI accident, the NRC isdeveloping new standards, revising existing standards,
and working with the national standards development
program to incorporate the lessons of TMI into national standards. In addition, the NRC Office ofStandards Development has assigned a number of its
senior staff to participate in the various TMI investigations and to assist in carrying out agency-wide activities during the post-TMI period. (See Chapter 2.)

Degraded Core Cooling. The TMI accident
involved a condition of inadequate cooling of the reactor fuel that led to substantial fuel damage andassociated release of radioactivity to the reactor
coolant and, into the containment. Accompanying thiscondition ot degraded core cooling was the generationof substantial quantities of hydrogen, attributable to
the reaction of the overheated fuel cladding with thereactor coolant. Furthermore, radioactivity was released into other parts of the plant and the surrounding environment. Degraded core cooling conditions in
accidents do not appear to be consistently handled in
NRC's regulations and guides. In view of this situation, a group has been established within the Office ofStandards Development to reevaluate the regulatory
requirements and guidance related to various designand operational aspects of the nuclear power plantsthat may be affected by degraded core cooling conditions that could occur during severe accidents. The objective of the group is to assess these regulations andregulatory guides in order to identify and implement
whatever improvements may be needed to better ensure that any systems and components that may be required to function during or following a degraded corecooling situation are designed, constructed, andoperated to do so.
Emergency Planning. A proposed rule change waspublished in the Federal Register in December 1979
that would require as a condition of operating license
issuance that State and local governmental emergency
response plans be submitted to and concurred in by the
NRC. In addition, during the 60-day public comment
period, several workshops will be held to (a) present
the proposed rule changes to State and local governments, utilities, and other interested parties, and (b)obtain comments concerning the costs, impacts, andpracticality of the proposed rule change. Commentsfrom the workshops and the public will be considered
prior to preparing a final rule in 1980.

198

REGULATIONS AND GUIDES
NRC standards are primarily of two types:• Regulations, setting forth in Title 10, Chapter I, of
the Code of Federal Regulations requirements thatmust be met.
• Regulatory Guides, describing, primarily, methods
acceptable to the NRC staff for implementing specificparts of the NRC's regulations.
When a new or amended regulation is proposed, it isnormally published in the Federal Register to allow in·terested citizens time for comment before final adoption,in accordance with the Administrative Procedure Act.
Following the public comment period, proposed regula~tions are revised, as needed, to reflect the comments
received. If the regulation is adopted by the NRC, it ispublished in the Federal Register in final form with thedate it becomes effective. After that publication, rules arecodified for inclusion in the annual publication of theCode of Federal Regulations.
Some regulatory guides delineate techniques used by thestaff to evaluate specific situations. Others provide
guidance to applicants concerning the information neededby the staff in its :review of applications for permits andlicenses. Many NRC guides refer to or endorse consensus
standards (alsQ called "national standards") that aredeveloped by reCognized national organizations, oftenwith NRC participation. NRC makes use of a national
standard in the regulatory process only after an independ~
ent review of the standard has been made by the NRC staff
and after public comment on NRC's planned use of thestandard has been reviewed.
The NRC encourages comments and suggestions for improvements in regulatory guides at all times, and they arerevised to take account of appropriate comments and suggestions and to reflect new information or experience. In
its continuing effort to provide for increased public participation in the regulatory process, the NRC instituted a
new procedure for developing and issuing guides during
this fiscal year. Guides are now. being issued for public'comment in draft form before complete staff review andbefore an official NRC staff position has been established.
Copies of draft regulatory guides, together with theirvalue/impact statements, are mailed for comment to many
individuals and organizations. The value/impact statement indicates the objective of the guide, its expected efft-'Ctiveness compared to alternative ways of achieving theobjective, and expected impacts on other safety systems,
NRC operations, other Government agencies, industry,
and the public.
In order to reduce the .burden on the taxpayer, the NRC
has made arrangements with the U.S. Government Printing Office to become a consigned sales agent for certain
NRC publications. Effective November 1, 1979,
regulatory guides are being included in this sales program.
Draft guides, which are issued for public comment, willcontinue to receive free distribution. Active guides will besold on a subscription or individual copy basis. Licenseesof the NRC will receive, at no cost, pertinent draft and ac.
tive guides as they are issued.
Proposed and efft-'Ctive regulations published during
fiscal year 1979 arc summarized in Appendix 4. Draft andactive regulatory guides issued, revised, or withdrawn arelisted in Appendix 5.

Decommissioning. NRC policy is being reevaluatedin this area with a view toward improving standards
for all nuclear facilities. Major technical studies arenearing completion on the engineering methodology,
radiation risks, and estimated costs of decommission~
ing light water reactors and other nuclear facilities. A
draft generic environmental impact statement (GElS),
to be used in developing appropriate regulations on
decommissioning of nuclear facilities, is also nearing
completion.
Spent Fuel Storage. Revision of proposed licens~ing requirements for independent spent fuel storage
installations is underway to reflect comments received
from the public and from the NRC staff. (See Chapter
4.)

Uranium Milling and Processing. Proposed and ef~
fective rule changes were published in August 1979 toestablish specific uranium mill licensing requirements.
These amendments are derived from a draft GElS on
uranium milling and the requirements contained in
the Uranium Mill Tailings Radiation Control Act of1978.
Inservice Inspection. The NRC is seeking to upgrade
the capability of inservice reactor inspection methods
to reliably detect and characterize flaws in components of the primary coolant and other safetyrelated systems. Contract studies are continuing, and aguide on inspection of welds in austenitic piping isunder development. A draft gUide has been issued on
inspection of welds in pressure vessels.
Siting Policy. In August 1979, a Siting Policy Task
Force, established by the NRC in November 1978,
recommended (NUREG-0625) extensive revision of
NRC's Reactor Site Criteria (10 CFR Part 100).
Similarly, Congressional committees and the President's Commission have recommended amendment of
the siting criteria, especially with regard todemographic factors such as population density .and
distribution around nuclear power plants. Action
pl~ns are being developed to implement the' recommendations.
Nuclear Power Plant Simulation. Concern for the
improvement of operator training led to the initiationof contract work with Oak Ridge National Laboratory
and Memphis State University Center for Nuclear
Studies to investigate the capabilities of nuclear power
plant simulators within the United States. In particular, the study is investigating the feasibility of increasing the use of simulators and expanding their rolein operator training. A regulatory guide to provide
gUidance concerning the use of simulators in operator
training is planned for the near future.
Improvement of Regulatory Guidance. As part of
the NRC's evaluation of the TMI accident, review of

199

regulatory guides and rules has been initiated to determine necessary improvements in regulatory guidance.
For example, Guide 1.B, on the qualifications of
nuclear power plant personnel, and Guide 1.33, on
overall quality assurance program requirements forthe operational phase of nuclear power plants, werereviewed to determine areas where guidance could be. improved. A notice for each of the guides was published in the Federal Register requesting specific comments that would improve the recommendations of theguides in these areas. In addition, work was initiatedto improve the regulatory guidance presented in Guide
1.97 on instrumentation to assess plant and environs
conditions during and following an accident at anuclear power plant.
Radiological Health. Major NRC efforts related tothe effects of low-level ionizing radiation included:(1) A joint EPA/NRC report was sent to the Congress on the research needs, capabilities, andcurrent programs of the two agencies withregard to health effects of ionizing radiation.

Standards development work initiated as the result of TMI included the preparation of new NRC guidance on the use of nuclear
power plant simulators and a study on simulator capabilities across(2) A contract was awarded for the preparation of astudy to identify and analyze the feasibility ofoptions for Federal epidemiological studies of
populations exposed to low-level ionizing radiation.
(3) NRC staff participated in preparing the reportof HEW's (now the Department of Health andHuman Services [HHS]) Interagency Task Forceon the Health Effects of Ionizing Radiation,issued for comment in April and in final form in
June.
(4) A Federal Interagency Task Force on Ionizing
Radiation Research was formed by HEW (now
HHS) with NRC participation. The NRC staff
also participated as members of a Task Forcesubcommittee formed to evaluate and makerecommendations on possible followup studiesof the residents in the TMI area.
(5) NRC staff members assisted HEW (now HHS)in designing a questionnaire for a population
census in the vicinity of TMI. This effort is ex-

the nation. This PWR control room simulator is in use at Consolidated Edison Company's Indian Point plant in New York.

200

safety of workers, the general public, and patients, butwith minimal intrusion into medical judgments affecting patients. Efforts in the regulation of radiopharmacies and on licensee reports of misadministrations inthis area are progressing, but these matters remainunresol ved.
Occupational Radiation Protection. The NRC isconsidering rule changes to strengthen its requirements that workers' exposures to radiation be
kept not only within regulatory limits but "as low as isreasonably achievable" (ALARA) and to make themmore readily subject to inspection and enforcement. In
addition, the NRC expects to participate in a public
hearing jointly with the EPA and the Occupational
Safety and Health Administration (OSHA) on new
EPA occupational exposure guidance for Federal
agencies, the adequacy of current occupational exposure standards, and related matters early in 1980.
Transportation. Sandia Laboratories continued
work on assessing the environmental impacts resulting
from the transportation of radioactive materials
through urban areas. Results concerning the consequences of the sabotage of spent fuel shipments influenced the decision to establish interim requirements
for the protection of spent fuel in transit. (SeeCh~pters 4 and 5.)

A health physics technician prepares to conduct a contamination
survey. The filter paper in her right hand is used to pick up surface dust or other material that may be contaminated. The portable survey meter carried over her shoulder is used to locate andidentify sources of radioactivity. Contamination surveys are conducted frequently in facili~ies where fuel is handled as well as at
power reactor plants.

pected to facilitate any future health effects
studies that may be performed on this population.
One of the major tasks of the Interagency Task
Force on the Health Effects of Ionizing Radiation was
to reexamine the organization of Federal radiation
protection programs. NRC staff members assisted that
Task Force in preparing recommendations to the
President on reorganization of Federal radiation protection and radiation research activities. Congressional hearings on this topic were held in May by asubcommittee of the Senate Committee on Governmental Affairs.
Nuclear Medicine. A final policy statement and rulechanges provide for NRC regulation of the radiation

Safeguards. Major staff efforts continue to befocused on (1) developing through regulations andguides a material control capability that is both timely
and sensitive with respect to material loss, and (2)implementing a material access authorization program for licensees. Other matters of importance include determining the level of safeguards needed toprotect SNM in the cores of non power reactors andpublishing the final rule to implement the US/IAEA
Agreement. (See Chapter 5.)
Radiation Protection Standards. The NRC has
begun to update and restructure its radiation protection standards as contained in 10 CFR Part 20. Publiccomments are being sought on areas of the present
regulations that need improvement.

POWER REACTOR STANDARDSDevelopment of power reactor standards continued
during fiscal year 1979, aimed primarily at protecting
the health and safety of the public and secondarily at
reducing the regulatory burden.

Surveillance and Inservice Inspection
Increased emphasis in this area brought about
several changes to regulations and the issuance of two
draft guides.
Section 50.55a, "Codes and Standards," of 10 CFRPart 50 has been amended to incorporate, by

==================================================201

0:

~
:O-®
I

10

1.
2.
3.
4.
5.
6.
7.

Bumper & Grill Area
8.
Engine Area
9.
Tire & Wheel Assemblies 10.
Cab Area
11.
Baggage Compartment 12.
Cab Sleeping Area
13.
Battery Boxes
14.

~l

@

I
I

..

Fifth Wheel Area
Trailer Refrigeration Unit
Ice Bunker Compartment
Roof~Both Tractor & Trailer
Under Entire Tractor~Trailer
Spare Part & Chain Compartments
Inte'rior of Trailer

15.
16.
17.
18.
19.
20.
21.
22.

Wheel Axles
Company Sign Panels
Document Pouches
Bumper on Rear of Trailer
Trailer-Upward Sliding Door
Light Lenses & Reflectors
Externally Mounted Air Filter
External Tractor Air Inlets

The potential for concealing special nuclear material in vehicles is
considered in developing safeguards standards. The diagram shown

above identifies 22 places for possible concealment.

reference, the 1977 Edition of the ASME Boiler andPressure Vessel Code, Division 1 of Section XI, "Rules
for Inservice Inspection of Nuclear Power Plants,"
with certain modifications, and Division 1 of Section
III, "Nuclear Power Plant Components," as well as
their addenda through 1978. This will result in more
flexibility for inservice inspection of pipe welds infacilities under construction and in operation and will
avoid potential conflict between the code and thetechnical specifications concerning examination requirements for steam generator tubing. This regulation was also amended to clarify certain ambiguities in
the requirements for inservice inspection of nuclear
power plants.
Appendix G, "Fracture Toughness Requirements,"
and Appendix H, "Reactor Vessel Material
Surveillance Program Requirements," to 10 CFR Part
50 were revised with regard to material toughness requirements for bolts and requirements for location andmethod of attachment of surveillance capsules.
Two draft guides were issued for public comment.
The first, issued in May 1979, deals with ultrasonic
testing of welds for inservice inspection and defines
some ultrasonic testing criteria considered acceptable

by NRC; the second, issued in August 1979, pertains toinservice inspection code case acceptability for ASME
Section XI, Division 1, and lists the ASME inspection
code cases that the NRC accepts.

Accident Analysis
The NRC is conSidering modifying the Emergency
Core Cooling System (ECCS) Rule (Section 50.46 of 10
CFR Part 50 and Appendix K to 10 CFR Part 50) totake into account experience derived from using therule in the licensing process, new research information, and reactor operating experience gained since therule was implemented. In December 1978, the NRC
published in the Federal Register an advance notice ofproposed rulemaking action and invited public comments and recommendations. The NRC is currently
evaluating the public comments as well as the lessons
learned from the Three Mile Island accident withrespect to the proposed ECCS rulemaking.

Reactor Containment
Containment Design. In October 1978, the NRC
published a regulation that is expected to Significantly

202

reduce the number of plants required to have inert
containment atmospheres in order to preventhydrogen explosions under certain accident conditions. This change takes account of increased conservatism in the revised emergency core cooling system
requirements. Revision 2 to Guide 1.7, which
describes acceptable methods of implementing the
new rule, was issued in December 1978. The NRC iscurrently reevaluating the regulation in view of the accident at Three Mile Island.
Concrete Containment and Structures. NRC endorsement of the ASME Boiler and Pressure Vessel
Code's Section III, Division 2, "Code for Concrete
Reactor Vessels' and Containments," progressed
another step with the issuance of Revision 1 in October
1978 and proposed Revision 2 in November 1979 to
Guide 1.136 on the subject of materials, construction,and testing of concrete containments. Acceptance of
this industry standard will make it possible towithdraw some existing regulatory guides to becovered by this standard.
In concrete containment buildings with prestressing
tendons that use a grease-like coating to inhibit corrosion, periodic inspections of these tendons are required. Based on knowledge gained from these testing
programs, Guide 1.35, on inservice inspection ofungrouted tendons in prestressed containments, is occasionally revised and updated. Proposed Revision 3,
reflecting an increased flexibility in the testing program and a broadening of its scope, was issued in April
1979. Guide 1.35.1, on determining prestressing forces
for inspection of prestressed concrete containment,
was simultaneously issued for comment to clarify the
NRC staff position on how to determine prestressing
forces when conducting the inservice inspection program outlined in Guide 1.35.
w

System and Component Criteria
General Design Guidance. A number of documents
have been issued in this category, including Revisions
14 and 15 to Guides 1.84 and 1.85. These guides continue to list acceptable ASME Boiler and Pressure
Vessel Code, Section III, Division 1, code cases as they
are published. NUREG-0554, "Single-Failure-Proof
Cranes for Nuclear Power Plants," was issued inAugust 1979 when Guide 1.104, on overhead cranehandling systems for nuclear power plants, waswithdrawn. NUREG-0554 identifies features of thedesign, fabrication, installation, inspection, testing,and operating of single-failure-proof overhead cranehandling systems that are used for handling criticalloads.
Guidance on Specific Systems and Components.
Guidance for the use of spray pond piping made fromfiberglass-reinforced thermosetting resin was updatedwith the issuance of Revision 2 to Guide 1.72. Also
issued was Guide 1.130, covering acceptable levels ofservice limits and appropriate combinations ofloadings for Class 1 plate-and-shell-type component
supports.
Revision 3 to Guide 1.70, providing the standard
format and content of safety analysis reports for lightwater-cooled nuclear power plants, was issued in
November 1978.
The following guide revisions were issued in October 1979 to reflect public comments: Revision 1 to
Guide 1.140, on design, testing, and maintenancecriteria for normal ventilation exhaust system air
filtration and adsorption units; Revision 1 to Guide
1.143, on design gUidance for radioactive waste
management systems, structures, and components;
and Revision 1 to Guide 1.137, which describes an acceptable method for complying with NRC regulations
regarding fuel-oil systems for standby diesel generators
and assurance of adequate fuel-oil quality.
A draft guide on the subject of safety-relatedpermanent dewatering systems was issued in September
1979.
A draft guide describing acceptable practices forcomplying with the NRC regulations with regard to
the nuclear analysis and design of concrete radiation
shielding for nuclear power plants was issued in
February 1979.
Several technical reports developing a decommissioning information base for light water reactors were
nearing completion at the end of the fiscal year. (See"Fuel Cycle Plant Standards" later in this chapter.)

Quality Assurance
Qu~lity assurance requirements for the design, constructIOn, and operation of safety-related structures,
systems, and components of nuclear power plants are
established in Appendix B to 10 CFR Part 50. During
the past fiscal year, the NRC issued new and revised
guides concerning the implementation of these requirements. In January 1979, Guide 1.144, on
auditing of quality assurance programs for nuclear
power plants, was issued for comment. In February
1979, Revision 2 to Guide 1.28, on ,quality assurance
program requirements for the design and constructionof nuclear power plants, and a draft guide on thequalification of audit personnel in the quality
assurance program for nuclear power plants were
issued. Also in February 1979, proposed Revision 2 to
Guide 1.8, on the qualifications of nuclear power
plant personnel, was issued and, following the Three
Mile Island accident, the public comment period was
extended to August 1979 to encourage additional
public input. The NRC requested specific comments inthe following areas covered by Guide 1.8:

203(1) Staffing, training, initial qualification, and requalification of operating personnel, supervisory personnel, and technical support personnel.
(2) Use of plant simulators for training, initialqualification, and requalification.
(3) Training of plant operating staff following extended shutdown.
(4) Content of programs for training nuclear power
plant personnel.
(5) Use of operating experience information intraining nuclear power plant personnel.
In July 1979, proposed Revision 1 to Guide 1.58, on
the qualification of inspection, examination, andtesting personnel for nuclear power plants, was issued.
Proposed Revision 3 to Guide 1.33, on overall quality
assurance program requirements for the operational
phase of nuclear power plants, was issued in August
1979. The revision to Guide 1.33 was initiated prior to
the Three Mile Island accident and provided improved
guidance that was developed as a result of a programusing feedback from the Office of Inspection and Enforcement to the Office of Standards Development.
Specific comments were requested in certain areas, inaddition to general comments, which would improve
the recommendations of Guide 1.33 in light of theThree Mile Island accident. These particular areas are
as follows:(1) Equipment control during the operational
phase, including procedural controls for conducting surveillance tests and inspections.
(2) Procedures for shift and relief turnover at
operating nuclear power plants.
(3) Quality assurance for maintenance, repair, andmodification activities at nuclear power plants.
(4) Preparation of operating and emergency procedures.
(5) Independence of operating plant personnel performing inspections of operating activities.
In September 1979, proposed Revision 2 to Guide
1.94, on quality assurance requirements for the installation, inspection, and testing of structural concrete and structural steel, soils, and foundations during the construction phase of nuclear power plants,
was issued.

Qualification TestingElectrical. Work continued on the development ofstandards and guides for the qualification testing ofelectric equipment used in nuclear power plants. A
draft guide on qualification testing of cable penetration fire stops was issued in July 1979. Proposed Revi-

sion 1 to Guide 1.131, on the qualification testing ofelectric cables and field splices, was issued in August
1979.
Supporting research continues at Sandia
Laboratories on test source equivalence, synergistic ef·
fects in environmental qualification, and accelerated
aging. Underwriters Laboratories completed the
NRC-sponsored study of the adequacy of IEEE Standard 383-1979 on flammability testing.
The NRC staff continued to participate with national standards committees in developing standards
for qualifying specific electric components that are important to safety-including converters, batteries,
battery chargers, inverters, motor control centers,
electrical connectors, and switchgear-as well as a
general standard for qualifying both electric andmechanical equipment. NRC also participated in the
updating of existing national qualification standards,
including those for qualifying valve operators,
penetration assemblies, continuous duty motors, andcables.
Mechanical. In February 1979, the NRC issued adraft guide on the subject of functional specification
for safety-related valve assemblies in nuclear power
plants. It describes a method for complying with requirements of Section III, "Design Control," of Appendix B to 10 CFR Part 50 related to the specifications for valve assemblies whose operability must be
ensured.

Electric Systems and Components
General Design Criterion 17, "Electric Power
Systems," of Appendix A to 10 CFR Part 50 includes arequirement that the onsite electric power system havesufficient capacity and capability to ensure that (1)specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are
not exceeded as a result of anticipated operational occurrences, and (2) the core is cooled and containment
integrity and other vital functions are maintained in
the event of postulated accidents. In November 1978,
Revision 1 to Guide 1.9, related to the design andtesting of diesel· generator units intended for use as onsite power sources in nuclear power plants, was issued
for comment. A draft guide on the protection of electric systems. and equipment from the effects of lightning was issued in August 1979.
The NRC staff continued to participate on national
standards committees in developing criteria for electric systems and components important to safety, including the updating of standards pertaining to protection system design, emergency power systemdesign, physical independence of redundant systems
and equipment, and post-accident monitoring.

204======~~====~~=================================

Reporting Defects and N oncompliances

Medical Certification

A rule (10 CFR Part 21) requiring certain persons toreport to the NRC defects that could create a substantial safety hazard, or failures to comply with regulations relating to substantial safety hazards, became
fully effective in January 1978. Following implementation of the regulation, an unintended impact occurred as a result of an interpretation of the term "basic
component," as defined by the rule. Revision of therule was completed in October 1979 to relieve the conditions resulting from that interpretation.

Revision 1 to Guide 1.134, which describes acceptable methods for providing the information neededby the Commission for its evaluation of the medical
qualification of applicants for initial or renewal
operator or senior operator licenses, was issued in
March 1979.

Protection Against FireThe extended one-year public comment period for
Guide 1.120 ended in November 1978. This guidedescribes how to implement NRC's requirement thatthe probability and effects of fire must be minimized
through fire prevention, detection, and suppression. It
also provides guidelines for designing fire safety
features into nuclear power plants. Public comments
are being reviewed and resolved in conjunction withthe earlier suggestion of the Advisory Committee on
Reactor Safeguards that the staff consider a "dedicatedshutdown system" instead of some of the individual
fire protection items called for in the guide.
As part of the program to develop fire protection
guidance for nuclear power plants, Sandia
Laboratories, under NRC contract, has completed
studies to develop technical bases in four separate
areas of fire protection: Task I-Ventilation; Task
2-Fire Detection; Task 3-Fire Barriers; and Task
4- Fire Hazards Analysis.
The final report for Task 1, NUREG/CR-0636,"Nuclear Power Plant Fire Protection; Ventilation(Subsystems Study Task 1)," was issued in August
1979. The report analyzes the function of ventilation
during a fire emergency at a nuclear power plant, its
interrelationship with other plant systems, and its rolein fire control and extinguishing operations.
The final report for Task 2, NUREG/CR-0488,"Nuclear Power Plant Fire Protection: Fire Detection(Subsystems Study Task 2) ," was issued in March
1979. The report examines the adequacy of fire detection in the context of overall nuclear plant safety.
The final report for Task 3, NUREG/CR-0468,"Nuclear Power Plant Fire Protection: Fire Barriers(Subsystems Study Task 3)," was issued in September
1979. The report considers the adequacy of the threehour fire barrier requirement of Guide 1.120.
The final report for Task 4, NUREG/CR-0654,"N uclear Power Plant Fire Protection: Fire Hazards
Analysis (Subsystems Study Task 4)," was issued in
September 1979. The report considers the presently
available methods of conducting a fire hazards
analysis and suggests using a composite of the best portions of all methods considered.

FUEL CYCLE PLANT STANDARDSThe NRC devoted substantial effort during fiscal
year 1979 to the development of standards concerning
the safety and environmental impacts of fuel cycle
plants.

Decommissioning
Technical studies for the NRC are continuing at theBattelle Pacific Northwest Laboratories (PNL) to
develop a decommissioning information base for lightwater reactors and other nuclear facilities. This basewill be used in developing appropriate regulations andguides. The PNL reports, "Decommissioning of
Nuclear Facilities-An Annotated Bibliography"(NUREG/CR-0131), and "Technology, Safety andCosts of Decommissioning a Reference Small Mixed
Oxide Fuel Fabrication Plant" (NUREG/CR-0129),were published in October 1978 and in February 1979,
respectively. Two PNL reports, "Decommissioning
Commercial Nuclear Facilities: A Review and Analysis
of Current Regulations" (NUREG/CR-0671) and"Technology, Safety and Costs of Decommissioning a
Reference Pressurized Water Reactor Power Station-Addendum" (NUREG/CR-0130) , were published in August 1979. Other PNL reports,"Technology, Safety and Costs of Decommissioning a
Reference Low-Level Waste Burial Ground"(NUREG/CR-0570) and "Technology, Safety andCosts of Decommissioning a Reference Boiling Water
Reactor Power Station" (NUREG/CR-0672) , were
nearing completion at the end of the fiscal year.
These PNL reports are part of the comprehensive
reevaluation of NRC policy related to decommissioning nuclear facilities. The detailed plan and schedule
for this reevaluation are described in an NRC staff
report entitled "Plan for Reevaluation of NRC Policy
for Decommissioning of Nuclear Facilities" (Revision 1
to NUREGI CR-0436) , which was published in
December 1978. Another NRC report, "Draft GElS on
Decommissioning of Nuclear Facilities"(NUREG-0586), was nearing completion at the end of
the fiscal year.
Two regional workshops were held in September
1979 to review the technical reports completed during
the fiscal year with State officials and to discuss withthem proposed policy and rule changes. Other draft
NRC reports prepared for these meetings, and also as a===============================================205
part of the reevaluation, were "Assuring the
Availability of Funds for Decommissioning Nuclear
Facilities" (NUREG-0584), "Thoughts on Regulatory
Changes for Decommissioning" (NUREG-0590), and"Residual Activity Limits for Decommissioning"(NUREG-0613) .
In June 1979, the NRC denied a petition by the
Public Interest Research Group, et aI., to initiate
rulemaking now to implement a specific decommissioning funding plan in which nuclear power plant
operators post surety bonds to cover decommissioning
costs. To the extent that the petitioners' request asked
the NRC to reconsider the adequacy of its regulations
on decommissioning, their request was granted.
However, a decision as to the specific method or
methods for funding decommissioning was deferred
until the comprehensive reevaluation of NRC policy
related to the decommissioning of nuclear facilities iscompleted.

Nuclear Criticality Safety
The following guide revisions were issued to reflectpublic comments: Revision 1 to Guides 3.34 and 3.35,
on assumptions used for evaluating the potential
radiological consequences of accidental nuclear
criticality in a uranium fuel fabrication plant and in aplutonium processing and fuel fabrication plant,
respectively, in July 1979; Revision 1 to Guide 3.43, on
nuclear criticality safety in the storage of fission
materials, in April 1979.

Plant Safety
Guidance in developing emergency plans for fuelcycle facilities licensed under 10 CFR Parts 50 and 70
is provided by Revision 1 to Guide 3.42, issued in
September 1979. This guidance may be further revised
as a result of assessment of the lessons of Three Mile
Island.

Spent Fuel Storage
Staff work is nearing completion on revisions to the
proposed 10 CFR Part 72, "Licensing Requirements
for the Storage of Spent Fuel in an Independent Spent
Fuel Storage Installation," to reflect changes made in
response to comments from the public and the staff .
.Guide 3.44, prOViding the standard format and content for the safety analysis report to be included in alicense application for the storage of spent fuel in an
independent spent fuel storage installation (waterbasin type), was issued for comment in December
1978. (See Chapter 4.)

Uranium Milling and Processing
Certain definitive rule changes to the Commission's
regulations have been made to establish specificuranium mill licensing requirements, particularly
with regard to the tailings or wastes generated during
the milling process. These rule changes incorporate into the Commission's regulations the conclusions derived from a generic environmental impact statement .
(GElS) on uranium milling and the licensing requirements covering "byproduct material" set forth in
the Uranium Mill Tailings Radiation Control Act(UMTRCA) of 1978. In August 1979, the NRC issued
effective rule changes to 10 CFR Part 40 and proposed
rule changes to 10 CFR Part 170 to establish specificlicensing requirements and fees for uranium mills andmill tailings (Le., byproduct material as defined by the
recently enacted UMTRCA of 1978), amendments to
10 CFR Parts 30 and 70 to require, for sake of consistency, the completion of a final environmental impact asessment prior to commencement of construction of certain types of major plants, and amendments
to 10 CFR Part 150 related to byproduct materials inAgreement States.

Standards for the decommissioning of fuel cycle facilities specify exhaustive inspections and ·tests or-buildings, grounds and equipment.
Here, an NRC insl'ector checks for radioactivity in concrete at adefunct fuel cycle facility in Chicago, Ill.

206

In order to define an acceptable detailed inservice
inspection program for earth and rock-fill embankments used to retain uranium mill tailings, Guide
3.11.1, on operational inspection and surveillance ofembankment retention systems for uranium mills, wasissued for comment in April 1979. This guide supplements existing Guide 3.11, on the design, construction, and inspection of embankment retention systems
for uranium mills.

Waste Management
During fiscal year 1979, the NRC Standards staff
participated in developing policy and proposed rules
for the licensing of high-level and low-level radioactive waste management facilities. A draft guide providing the standard format and content of license applications for the disposal of high-level radioactive
waste in geologic repositories was nearing completionat the end of the fiscal year.
In November 1978, the NRC published for commenta proposed general statement of policy suggesting
licensing procedures for geologic disposal of high-level
radioactive wastes. Concurrently, a draft of an implementing regulation was circulated to the States fortheir review. This was done as part of NRC's policy toseek early input from the States in the waste management area. Comments on the proposed policy statement and comments from the States have been considered. The technical section of a proposed regulation(10 CFR Part 60) for geologic disposal of high-level
waste is under preparation; the procedural section ofthis proposed regulation was issued for public comment in December 1979.
In the area of high-level waste, site suitability
criteria were developed for the Technical Requirements of the proposed 10 CFR Part 60. Work was
begun on branch technical positions on long-termclimatic change, assessment of geochemical retardation, hydrologiC assessment for transport modeling,
and the use of models in site evaluation. Future work
will emphasize specific requirements and guidance forsite characterization in safety and environmental
analysis reports.
A similar program of regulations and gUides isunderway to address the management of low-level
radioactive wastes. Ongoing efforts include a regulation for licensed disposal of low-level radioactive
wastes that addresses waste categorization. Several
regulatory guides are planned in support of thisregulation.

General Site Suitability Criteria
A study is being conducted under contract with
Argonne National Laboratory to identify andcharacterize various typical accidents and incidents
that have occurred at fuel cycle facilities. The identification of these accidents and their associated exposures of personnel to radioactivity will provide an
updated assessment of the impact at each type of
facility.
A study was completed under contract with United
Engineers and Constructors, Inc., to collect available
data and establish cost characteristics for nuclear
generating station designs, as they are determined by
the most significant site characteristics. The scope ofthe study included main condenser cooling, transmission, flood protection, site access, and plant/site earthwork.

SITING STANDARDSThe standards on the siting of nuclear plants deal
with procedures for site review, site safety, and protection of the environment.

Site Review Procedures
Alternative Sites. A proposed rule has been preparedto establish regulatory policy and procedures forevaluation of alternative sites for nuclear generating
stations under the National Environmental Policy Act(NEPA). The proposed rule is designed to (1) fulfill the
NEPA objectives of ensuring that environmental factors have been fully considered in NRC decisionmaking, (2) reduce uncertainty and delay in the decisionmaking process, (3) reduce Federal paperwork in
NEPA statements, and (4) limit alternative site review
to relevant and material issues. There is an ongoing
contract at Brookhaven National Laboratory to survey
the range of site screening and selection
methodologies, to examine the operational features,and to identify those features that should be looked forin determining whether an applicant has employed an
acceptable selection methodology.
Emergency Planning. Guide 2.6, on emergency
planning for research reactors, was issued for comment in January 1979. The staff participated in the
preparation and writing of NUREG-0396, EPA
520/1-78-016, entitled "Planning Basis for theDevelopment of State and Local Government
Radiological Emergency Response Plans in Support ofLight Water Nuclear Power Plants." This report of the
NRC/EPA Task Force on Emergency Planning wasissued for public comment in December 1978.
The NRC is in the process of reevaluating and revising regulations and regulatory guides in the emergency
planning area. Late in 1979, the Commission published proposed amendments to its regulations to provide an interim upgrade of NRC emergency planning
rules.
In June 1979, the NRC began a formal reconsideration of the role of emergency planning in assuring the================================================207
continued protection of the public health and safety inareas around nuclear power facilities. The Commission had begun this reconsideration in recognition of
the need for more effective emergency planning and in
response to reports issued by responsible offices of theFederal government and its Congressional oversightcommittees.
On July 17, 1979, the NRC published an advancenotice of proposed rulemaking on the subject of Stateand local emergency response plans and those oflicensees. Comments received from the advance notice
are being evaluated by the NRC staff for use in possible future rulemaking actions.
By memorandum dated July 21, 1979, the Commission requested that the NRC staff undertake expedited
rulemaking on the subject of emergency planning. The
proposed amendments (published in the Federal
Register in December 1979) respond to the Commission's request. Consequently, because these proposed
amendments were prepared on an expeditious basis,considerations related to the workability of the proposed rule changes may have been overlooked andsignificant impacts to NRC, applicants, licensees, andState and local governments may not have been identified. Therefore, the NRC will seek additional publiccomment by holding workshops prior to the preparation of a final rule to (a) present the proposed rulechanges to State and local governments, utilities, andother interested parties, and (b) to obtain comments
concerning the costs, impacts, and practicality of the
proposed rule changes.
Specifically, NRC is proposing amendments to 10
CFR Part 50 to require that emergency response planning considerations be extended to Emergency Planning Zones (discussed in NUREG-0396). Both the Commission and EPA have formally endorsed the concepts
in that EPA/NRC Report, 44 Federal Register 61123(October 23, 1979).
The proposed amendments include, as a conditionof operating license issuance, that State and local
governmental emergency response plans be submitted
to and concurred in by the NRC. If the State and local
governmental plans have not received NRC concurrence, the applicant may attempt to demonstrate to
the satisfaction of the Commission that deficiencies in
the plans are not significant for the plant in question
or that alternative compensating actions have been or
will be taken.
In addition, the proposed rulemaking revises 10
CFR Part 50, Appendix E, "Emergency Plans for Production and Utilization Facilities," to clarify, expand,and upgrade the Commission's emergency planning
regulations.
The Commission is also proposing to amend its
regulations in order to require certain special nuclear
material facility licensees (for processing and fuelfabrication, scrap recovery, or conversion of uranium
hexafluoride) to keep their emergency plans up to date.

Site Safety
NRC site safety standards are rules and guides forassessing and mitigating adverse effects associatedwith natural events such as earthquakes, floods, andextreme meteorological conditions and man's activities
at and near nuclear sites.
In the field of meteorology, Guide 1.145, on atmospheric dispersion models for potential accident
consequence assessments at nuclear power plants, wasissued for public comment in August 1979.
NUREG/CR-1024, "An Initial Assessment of Flash
Density and Peak Current Characteristics of Lightning
Flashes to Ground in South Florida," was issued in October 1979. A revision to Guide l.23, on onsite
meteorological programs, is in progress. The staff iscontinuing data evaluation for the development ofstandards on extreme windspeeds in coastal areas, extreme snow and ice accumulations, and extreme
temperatures, as well as on the hazards associatedwith lightning and dust and sand storms.
In the geology and seismology area, review of Appendix A, "Seismic and Geologic Siting Criteria for
Nuclear Power Plants," to 10 CFR Part 100 was completed. The results of the staffs review were provided
to the Commission on April 27 , 1979. Other earthscience activities included gathering data for atechnical report on dating materials that may be foundin fault zones and classifying the faults and fractures of
the Appalachian foldbelt. A study on siting nuclear
facilities in areas susceptible to ground collapse is still
underway.
In the hydrology area, work on a revision to Guide
1.135, on normal water level and discharge, is in progress. ANSI committee and IAEA work on surface
water, ground water, and radion~clide transport isalso continuing through the preparation of new standards and revision of existing standards and guides. Input was provided to the reassessment effort on lowlevel waste regulations, and work is continuing on thehydrologic aspects of siting criteria.

ENVIRONMENTAL PROTECTION
Environmental protection standards are concernedwith the protection of the public and the environment
from both radiolOgical and nonradiological impacts ofnuclear facilities. This includes assessment of environmental impacts, control of effluents, andmonitoring of the environment around the facilities.
Revision 1 to Guide 4.15, on quality assurance forradiological effluent and environmental monitoring
for normal operations, was issued in February 1979.
An NRC Task Group issued its report on "Radiological
Monitoring by NRC Licensees for Routine Operationsof Nuclear Facilities," NUREG-0475, in October
1978.

208

A regulation was proposed to revoke Section 20.304
of 10 CFR Part 20, which currently allows licensees tobury small quantities of radionuclides without notifying NRC.
A substantial effort is now being devoted to the environmental aspects of uranium milling, decommissioning and decontaminating nuclear facilities, andradioactive waste disposal.

Interagency Coordination
NRC has the responsibility for implementing both
EPA's guidance and generally applicable environmental standards for protection against radiation.
During 1977 , EPA published standards (40 CFR Part
190) that limit releases of radioactive material andresulting doses to the public from the operation ofvarious nuclear facilities associated with the uranium
fuel cycle. An NRC task force, which includes EPA
staff members; is completing the program for
implementing these standards.
The Clean Air Act Amendments of 1977 include
provisions for EPA to develop emission standards forradioactive materials from NRC-licensed facilities.
EPA is expected to issue its first standard in 1980. The
NRC and EPA staffs have been working together on
the development of these standards and on an interagency agreement to minimize duplication of effort in
implementing the Clean Air Act.
The NRC became a member, along with 15 othermajor Federal agencies, of the Toxic Substances
Strategy Committee, formed under the leadership of
the Council on Environmental Quality. NRC staff
members served on seven of the task groups of thiscommittee, which is to recommend strategies to be
used by the Federal Government for the control of toxic and hazardous substances. Although radioactive
materials have been excluded from this report, theprinciples for controlling cancer-causing materials
would be expected to affect radiation control
strategies, and the expertise gained by NRC in controlling radiation is directly applicable to some aspectsof controlling other carcinogens. The draft report of
the Committee has been issued for public comment.
The final version is expected to be transmitted to the
President by the end of the year.
The NRC staff has reviewed the proposed Hazardous Waste Guidelines and Regulations issued by the
EPA Division of Hazardous Waste, and a series ofmeetings has been held with the EPA staff with a view
to clarifying apparent differences in approach fordisposal of EPA-regulated hazardous wastes and NRCregulated low-level radioactive wastes. A memorandum of understanding is being developed to provide
guidance for those instances when EPA-regulated
hazardous wastes are mixed with radioactive wastes.
The Departments of Interior and Commerce havepu~lished proposed regulations implementing the Fish

and Wildlife Coordination Act. These proposed
regulations could require significant revisions in
NRC's operating procedures. The NRC staff has commented on the proposed regulations and, dependingon the final form of the regulations, will develop implementing procedures consistent with the NRC's position as an independent regulatory agency. Continuing
discussions are expected among the three agencies todetermine how this can be done in a mutually acceptable manner.

RADIOLOGICAL HEALTH STANDARDS
Low*Level Radiation Effects
Significant progress was made during the year incarrying out the mandates of Public Law 95*601,
which directed the NRC and the EPA, in consultation
with the Secretary of Health, Education and Welfare(HEW, now HHS), to conduct preliminary planningand design studies for epidemiological studies. A joint
EPA/NRC report on the health effects research needs,
capabilities, and current programs of the two agencies
was prepared and transmitted to the Congress. A contract was awarded for the preparation of a plan for
identifying the options for Federal epidemiologicalresearch and for assessing the feasibility of these options. A progress report to the Congress on these activities was prepared and transmitted at the end of the
fiscal year.
The report of the Interagency Task Force on the
Health Effects of Ionizing Radiation was issued by
HEW. Drafts of the various work group reports were
issued for public comment in April, and the final
report was issued in June. The task force examined information on the health effects of low-level ionizing
radiation and the problem of coordination of Federal
radiation protection and radiation research activities.
The Director of the Office of Standards Development
represented NRC on the task force, and NRC staff
members participated actively on the various working
groups. The report was the subject of hearings held on
May 8, 1979, before the Subcommittee on Energy,
Nuclear Proliferation, and Federal Services of the
Senate Committee on Governmental Affairs. The
NRC was represented at this hearing and contributed
suggestions for improving Federal radiation protection
and radiation research programs.
NRC staff members were also active participants inthe efforts of the Interagency Task Force on Ionizing
Radiation Research to improve the coordination ofFederal research programs, to conduct a comprehensive review of these programs, and to establish a comprehensive program of research into the biological effects of low-level ionizing radiation. The task force
was convened by the Secretary of Health, Education,and Welfare to assist in carrying out the requirementsof Title II of Public Law 95-622, the Biomedical

==================================================209
Research and Research Training Amendments of 1978.
A special subcommittee of this task force was formed
after the accident at the Three Mile Island Nuclear
Station to evaluate proposed studies and to recommend possible Federal studies of the health andpsychological impact of the accident. NRC staff weremembers of this subcommittee and provided information on the radiation levels and doses received by offsite residents and workers at the facility.
A final report was published on an NRC-sponsored
epidemiological study of residents in Mesa County,
Colorado, where extensive uranium mining and milling operations are located. This study, conducted by
the Colorado Department of Public Health, found no
relationship between exposure to mill tailings used inresidential construction and an observed excess ofleukemia in Mesa County.
Three contracts for independent analyses of data on
the radiation exposures and their effects on the healthof the Hanford worker population were also completed. The results of this work will be useful in support of the epidemiology feasibility planning studies
described above. Valuable insights were obtained into
the complexities of handling and analyzing data basesof this nature in investigating possible causal relationships between radiation exposures and health effects.

In April 1979 the NRC Office of Nuclear Reactor Regulation
began a public whole body countingrrogram at Three Mile Islandusing this "do-it-yourselr' counter. 0 more than 700 counts

completed none identified radioactivity which could have resulted
trom the accident. The radiation detector, housed in the box-likestructure, passes over the examinee, on signal.

Nuclear Medicine
In fiscal year 1979, the NRC issued a final medical
policy statement, a final rule requiring radiation
surveys of certain therapy patients prior to discharge,
and a final rule requiring annual calibrations ofteletherapy units. The two rules resulted from reportsof medical misadministrations of radiopharmaceuticals. The NRC also issued a proposed rule requiring each medical licensee to appoint a radiation

210

safety committee and one requiring medical licensees
to test certain radiopharmaceuticals for radionuclide
contaminants. A final rule change, effective in June
1979, requires all materials licensees, including
medical institutions, to inform the NRC when theydecide to give up their licenses. Another final rulechange, effective in September 1979, added
veterinarians to the groups authorized to use
byproduct material under general license for in vitro
clinical or laboratory testing.

OCCUPATIONAL HEALTH STANDARDS
Respiratory Protection
During fiscal year 1979, the NRC continued todevelop and provide the information necessary to ensure the adequacy of licensees' respiratory protection
programs. Two Respirator Users' Warnings from the
National Institute for Occupational Safety and Health(NIOSH) were circulated to advise licensees of actions
to take concerning potential failures of certain selfcontained breathing apparatus (SCBA) (Inspection
and Enforcement Circulars 79-09 and 79-15).
Under a technical assistance contract with the Los
Alamos Scientific Laboratory (LASL), a videotape wasproduced for training users in proper respirator fitting
methods. Under a separate research contract with
LASL, measurements were made to determine theprotection provided by open-circuit SCBA. These
data, along with additional information from LASL
and other sources, will be used to update the NRC's
guidance to licensees on acceptable respiratory protecticm programs.
LASL is also providing technical assistance for the
NRC's Three Mile Island post-accident operations. An
expert on respiratory protection provided onsite
assistance to the NRC. Special studies were carried out
at the laboratory to obtain information on the performance of new positive-pressure closed-circuit SCBA
devices and on the efficacy of air-purifying respirator
canisters for use against airborne radioiodines.
The NRC continued its cooperation with othergovernmental and nongovernmental agencies toward
the development of needed improvements in
respiratory protection.

Testing for Personnel Dosimetry
Evaluations of the degree of accuracy that is provided by personnel dosimetry processors in the UnitedStates indicate that improved performance of someprocessors is needed. Personnel dosimetry devices areused to measure the radiation dose received by workers
in NRC-licensed facilities. To obtain more accurate
processing of dosimeters, the NRC staff is working on arequirement that personnel dosimetry results be ac-

cepted only from a processor who has successfully
passed certain prescribed accuracy tests. The test
criteria would be adapted from a consensus standard
developed by ANSI.
In preparation for the new regulation, the NRC
funded a two-year pilot study, which was conductedby the University of Michigan. The objectives of thepilot study were:(1) To test the consensus standard for practicality as
well as for degree of difficulty.
(2) To provide processors an opportunity to correct
any process problems that they may have prior
to publication of the new regulation in effective
form.
(3) To develop a detailed procedures manual for use
by future testing laboratories.
The study was completed in September 1979. Fiftyeight processors participated in two rounds of testingin various radiation categories. In Test # 1, only 22 percent of the category tests were passed, using criteria
published in the ANSI standard (NI3.11, July 1978).
In Test #2, the percentage of category tests passed roseto 38 percent. Thus, while considerable improvement
was experienced, the overall performance of some processors is evidently in need of improvement. The NRC
staff has initiated work on the necessary amendment of
10 CFR Part 20. The pilot study indicated that the
ANSI standard, with minor modifications, is suitable
for the regulatory purposes of the NRC.

Personnel Monitoring Reports
In September 1978, the NRC published an amendment to 10 CFR Part 20 to extend to all NRC licensees
the requirement for annual statistical summary reports
on workers' radiation exposures. Under the previous
regulation, only four categories of licensees were required to submit an annual statistical summary ofmonitored whole-body exposures, i.e., the number ofpeople in each of 18 prescribed ranges of radiation exposure.
The amendment to Part 20 extends this statisticalsummary reporting requirement to all NRC specificlicensees for a period of two years. After evaluating thedata for 1978 and 1979, the NRC will consider
whether it will extend or modify the reporting requirement. The four categories of licensees previously
covered will continue to be required to report in any
event. The amendment does not affect existing requirements for the provision and use of personnel
monitoring equipment or for the recording of personnel monitoring data but relateS solely to the reportingof data already recorded.

===================================================211
Worker Exposure to Neutrons
The use of NTA film has come into question as anadequate dosimeter for monitoring workers' exposures
to neutrons at pressurized water reactors (PWRs). The
NRC has contracted with the Battelle Pacific Northwest Laboratories (PNL) and the Rensselaer
Polytechnic Institute (RPI) independently to examine
this problem. PNL and BPI are conducting surveys
both of the neutron spectra in work spaces at PWRs
and of the response of various types of dosimeters tosuch neutrons. The preliminary results of these studies
indicate that NT A film might not be a good dosimeterfor these neutrons because it responds poorly to theneutrons of less than 0.7 Mev that constitute a portionof the neutrons in PWR work spaces. Improved albedo
dosimeters might provide better measurements ofthese neutron exposures. Therefore, the NRC undertook a revision of Guide 8.14 in 1979. The purpose ofthis revision is to advise licensees of the inadequacy of
NTA film for neutron dosimetry for routine operationsat nuclear power plants. In revising the guide, the
NRC will address the "state of technology" in personnel dosimeters for neutrons and will provide guidanceon acceptable methods for estimating workers' exposures to neutrons.

Calibration of Air Sampling Instruments
In October 1979, a draft regulatory guide was issued
on acceptable methods of calibrating air sampling instruments to more accurately determine the volume of
air sampled. In addition, a frequency of calibration,
an error limit, and documentation requirements are
specified.
The gUide is expected to improve licensees' air
monitoring programs and estimates of workers' exposures to airborne radioactive material.

Transient Worker Radiation Protection
In June 1979, the NRC published amendments to 10
CFR Parts 19 and 20 that are intended to minimize thepossibility of overexposure of (1) short-term workers,
sometimes called "transient workers," and other individuals who may be employed by or work in the
restricted areas of more than one licensee within asingle calendar quarter, and (2) individuals who may
work for more than one licensee at a time
(moonlighters) .
The amendments, which became effective on
August 20, 1979, require NRC licensees to control the
total occupational radiation dose of individuals who
work in NRC-licensed activities. The changes require
licensees (1) to obtain information from a prospective
employee on occupationally related doses received
during a current calendar quarter from sources outside

This lOO·foot radon sampling tower near a mine in New Mexico
measures the speed, direction and stability of winds. temperatureand the atmospheric pressure, and takes ambient radon concentrations measurements at five different elevations. This program isconducted for NRC's Office of Nuclear Materials Safety and
Safeguards by Argonne National Laboratory.

the licensee's control if there is a Ghance. that theemployee may subsequently receive a dose in excess of
25 percent of the regulatory standards in the facility of
the new employer; (2) to furnish prompt estimates ofoccupational dose, at the request of the individual,
upon termination of work; and (3) to keep associatedrecords.

Medical Institutions
Two guides specific to occupational radiation protection in medical institutions were issued for comment during fiscal year 1979: Guide 8.23, on radiation
surveys in medical institutions, and Guide 10.8, on
medical licensing. Guide 10.8 explains the information
to be submitted in an application for a license to use
byproduct radioactive materials in diagnostic andtherapeutic medical applications, provides a simpler
form (NRC-313M) for completing the required entries,
and provides acceptable methods and statements
related to radiation safety and user qualifications.

212

As reported last year, Guide 8.18 and a companion
report (NUREG-0267) that provides more detailed information and references were issued for comment inJanuary 1978. During fiscal year 1979, these two
documents have provided broad interim guidance andinformation for establishing acceptable occupationalradiation safety programs in medical institutions. Both
documents have received a generally favorable reception from the medical and medical physics communities, but there have been a number of suggestions
for improvement, additions, or deletions that will require a careful balancing of the various viewpoints in
preparing the final versions. Extensive comments havealso been received on Guides 8.23 and 10.8. Since thefour documents now published for medical institution
radiation safety and licensing have some overlapping
subject areas of guidance, a major effort will be madeduring the early part of fiscal year 1980 to revise all ofthese documents. The effort will include considerationof the advice of interested professional and public
groups who may be interested in or affected by theguidance in these documents.

Bioassays
Revision 1 to Guide 8.20, providing guidance for
1-125 and 1-131 bioassay, was issued in September
1979. A draft guide on bioassay programs and methods
for fission and activation products was issued inAugust 1979. For the most part, this draft guide adopts
the recommendations of a recent standard issued by
the American National Standards Institute (ANSI).
The NRC staff participated in the development of the
ANSI standard. These documents supplement threeprevious guides that provided general information on
acceptable methods of interpreting bioassay results
and gave specific guidance on interpreting uranium
bioassays. The new documents provide guidance tomanagement on the levels of radioactivity or working
conditions under which bioassay should be performed.
They also specify on whom the assays should be performed and when investigative or corrective measures
should be taken. The iodine and fission productbioassay guidance takes into consideration theamounts of each radionuclide above which exposure
potential becomes appreciable, as indicated by industrial and medical experience.

Health Protection at Uranium Mills
A memorandum of understanding to ensure consistency of regulatory actions is still being developed
between the NRC and the Mine Safety and Health Administration (MSHA) of the Department of Labor.
The Federal Mine Safety and Health Act of 1977 gives
MSHA jurisdiction with respect to protection of
uranium mill workers similar to that given NRC underthe Atomic Energy Act of 1954, as amended.

Reports were published during this fiscal year on thesolubility of yellowcake in the human body
(NUREG/CR-0414 and NUREG/CR-0530) and on invivo counting for uranium and radium in the body
(NUREG/CR-0841) .

A regulatory guide on health physics surveys at
uranium mills is in preparation.
Occupational ALARA at Uranium Mills. In 1979,
the NRC undertook the development of a guide intended to provide operators of uranium recovery
facilities with the specific information needed to ensure that occupational exposures are as low asreasonably achievable (ALARA) for workers inuranium mills. In providing adequate occupational
health protection for uranium mill workers, it is important to consider the chemical toxicity of uranium tothe kidney, as well as radiologic effects. The major
health effect for uranium mill workers results from theinhalation of suspended particulates. Thus, theALARA guide will provide both operational procedures and design specifications for ventilation equipment to minimize suspended particulate materials inwork areas. Of particular concern are those areas
associated with the yellowcake dryer and with the
packaging and shipping of dried yellowcake. These
locations are important because yellowcake is abiologically soluble compound and has a provenchemotoxic effect on the kidney, as well as a radiotoxic
effect. The guide is scheduled for issuance for publiccomment by mid-1980.

Effecting Occupational ALARA
The NRC staff has completed consideration of
methods for quantifying ALARA requirements under
proposed amendments to NRC regulations previously
developed. These proposed amendments would require the development and implementation of ALARA
programs by all licensees who are required by the NRC
to perform personnel dosimetry, air sampling, orbioassays for worker protection. The staff is proposing
the development of several guides to complement the
proposed amendments. The guides would be specificto several different types of licensees and would provide descriptions of ALARA programs that would be
acceptable under the proposed amendments.
Surve~s During U-235 Processing andFuel Fabrication

GUi.de 8.24, issued for comment in November 1978,
identifies the types and frequencies of health physics
surveys that are acceptable to the NRC staff for use inplants licensed by the NRC for the processing ofenriched uranium-235 and for the fabrication of
uranium fuel. Revision 1 to Guide 8.24 was issued in
October 1979.

=================================================213
Radiation Safety Training at LWRs

SAFEGUARDS STANDARDS

A draft guide on radiation protection training for
personnel at light water reactor (L WR) facilities was
issued in August 1979. The draft guide describes radiation protection training consistent both with maintaining occupational doses at L WRs as low as is reasonably
achievable and with the requirement of 10 CFR Part
19 for training individuals who enter restricted areasat nuclear power plants.

The NRC devoted substantial standards effort during fiscal year 1979 to the safeguarding of nuclear
materials and facilities against theft and diversion.
Development of regulations in this area is discussed in
Chapter 5.

Gamma Irradiators

Physical Protection

A draft guide on the preparation of license applications for the use of gamma irradiators was issued in
March 1979. This guide informs people who wish to
apply for such licenses of the information that the
NRC requires for reviewing and acting upon such applications. The guide reflects the new requirements forirradiators in 10 CFR Part 20, effective in 1978.

In support of existing regulations and of the proposed and newly adopted safeguards regulations forphysical protection of SNM discussed in Chapter 5, the
NRC has issued several regulatory guides andtechnical reports. These include:

Industrial Radiography Safety
New amendments to 10 CFR Part 34 on safety in industrial radiography were published in effective formin September 1979.
A petition for rulemaking to have the NRC license
individual radiographers is under consideration. The
petition states that safety in industrial radiography
could be improved by making individual
radiographers more directly responsible for their actions.
A draft regulatory guide on the use of audible-alarm
dosimeters for improving radiography safety wasissued in August 1979. Battelle Pacific Northwest
Laboratories is conducting tests of such dosimeters todetermine their reliability, and a report on their firstseries of tests has been published (NUREG/CR-0554).

Petition on "Radiation Area"
A petition received in October 1978 requested that
the NRC amend the definition of "radiation area," as
set forth in §20.202 (b) (2) of 10 CFR Part 20, to specify
dose rates comparable to those set forth in §20.105,
rather than the rates of 5 millirems in an hour and 100
millirems in any 5 consecutive days which are currently specified in §20.202(b)(2). The proposed change
would require any area that could not qualify as anunrestricted area to be posted as a radiation area.
After due consideration of the petition and the bases
for the proposed change provided by the petitioner,
the comments received following publication of noticeof receipt of the petition, and other factors involved,
the petition was denied on September 26, 1979. The
principal reason for the denial was that there does notappear to be any reduction in risk associated with thepetitioned change. Indeed, there is a potential that unnecessary exposure of workers might result.

(1) Proposed Revision 1 to Guide 5.7, describing
measures acceptable to the NRC staff for implementing entry/exit control requirements at
facilities other than nuclear power plants, wasissued in May 1979.
(2) Proposed Revision 1 to Guide 5.14, describing
measures acceptable to the NRC staff forsurveillance or observation of individuals withinmaterial access areas in order to strengthen thesafeguarding of strategic special nuclear
material, was issued in May 1979.
(3) Proposed Revision 2 to Guide 5.44, describing six
types of perimeter alarm systems for detecting intrusion into plants that use or process highly
enriched uranium, uranium-233, or plutonium,
was issued in May 1979. This guide also sets forthcriteria that are acceptable to the NRC staff forthe systems' performance and use.
(4) Proposed Revision 1 to Guide 5.57, describing
procedures acceptable to the NRC staff withregard to the protection of strategic special
nuclear material during preparation for shipment, transfer between licensees, and on receipt
by a licensee, was issued in May 1979.
(5) A draft guide providing a standard format andcontent for licensees' plans submitted in responseto requirements for the protection of Category II
and III type materials against theft was issued inAugust 1979.
(6) A draft guide providing a logical scheme fordetermining when a safeguards event should bereported to the NRC, in addition to a partial listof the kinds of events that should be reported,
was issued in October 1979.
(7) NUREG/CR-0464, "Site Security Personnel
Training Manual," and NUREG/CR-0465,"Transportation Security Personnel Training

214

NRC safeguards regulations for fuel cycle facilities specify
many of the techniques and equipment to be used for theprevention of intrusion by unauthorized people. Some
devices used for this purpose are shown nere. Clockwisefro~ to,? left: a fenced, protected area features ni~ttime il.
lummation and infrared intrusion detection devices; closed-

circuit TV surveillance equipment scans protected areas; access control points feature shielded doors/windows, two-way
remote communications, internal controls for lighting
systems and gatelocks, as needed; a central alarm control
panel "oversees" an entire facility.

==================================================215
Manual," assist licensees in developing effec·
tive security personnel training and qualification programs, as required by 10 CFR Part 73.
The manuals identify the level and scope oftraining of security personnel assigned to per·form specific tasks and job duties to protect
special nuclear material, nuclear facilities,
and shipments.
(8) NUREG/CR-0484, "Vehicle Access and Control
Planning Document," provides guidance inplanning a vehicle access and control system atnuclear fixed site facilities.
(9) NUREG/CR-0485, "Vehicle Access and SearchTraining Manual," provides training guidance
to help NRC licensees better perform vehicle ac·cess and search operations at nuclear fixed sitefacilities.
(10) NUREG/CR-0509, "Emergency Power Supplies
for Physical Security Systems," includes basic
information useful in the planning, design, andimplementation of emergency electric powersystems for safeguard security systems.
(11) NUREG/CR-0510, "Duress Alarms for Nuclear
Fixed Site Facilities," identifies and discusses
different types of devices currently available
that could be used as duress alarm systems.
(12) NUREG/CR-0543, "Central Alarm Station andSecondary Alarm Station Planning Document,"
provides planning guidance for the siting, construction, and equipping of a central alarm station and a secondary alarm station as requiredby the NRC.

Material Control and Accounting
In support of existing requirements and thestrengthened regulations for material control and accounting of special nuclear material (SNM), discussed
in Chapter 5, the NRC issued several regulatory guides
and technical reports. These include:(1) Guide 5.58, issued for comment in November
1978, presents conditions and procedural approaches acceptable to the NRC staff forestablishing and maintaining traceability of
SNM control and accounting measurements.
(2) NUREG/CR-0061, "Preparation of Working
Calibration and Test Materials: Plutonium Oxide,;; provides guidance for preparing
plutonium dioxide reference materials used tocalibrate and to maintain quality control overmethods of analysis for plutonium.
(3) NUREG/CR-0139, "Preparation of Working
Calibration and Test Materials: Mixed Oxide,"
provides guidance for preparing mixed-oxidereference materials used to calibrate and tomaintain quality control over methods ofanalysis for plutonium and uranium in mixedoxide powders and pellets.
(4) NUREG/CR.0515, "Methods for the Accountability of Reprocessing Plant Dissolver andWaste Solutions," gives detailed methods for theaccountability of uranium and plutonium solutions resulting from the dissolution of spent lightwater reactor fuel.
(5) NUREG/CR-0562, "Specifications for Germanium Radiation Detectors Used for Gamma
Ray Assay in Safeguards Applications,"
discusses the physics and engineering of radiation detector fabrication and operation and thebasic principles of assay by gamma-ray spectroscopy.
(6) NUREG/CR.0591, "Current Usage of Containers for SNM Storage, Transfer and
Measurements," is an interim report thatsurveyed the types, sizes, and materials of containers used to store, transfer, or measure SNM.
Information gained from this report will serveas the basis for standardization of SNM containers with respect to size and fabrication
specifications in the nuclear industry, reducing
cost while improving nondestructive assay
measurement performance.
(7) NUREG/CR.0602, "Active Assay Handbook,"
contains principles of active nondestructive
assay and associated measurement procedures
for a variety of SNM samples.
(8) NUREG/CR-0683, "Statistical Methods for
Evaluating Sequential Material Balance Data,"
evaluates and compares four material loss
estimators for two different loss mechanisms
and for four different time periods using thecriterion of "power of the test" for detecting aloss.
(9) NUREG/CR-0772, "Auditing of Measurement
Control Programs," documents concepts andfactors associated with auditing a measurement
control system that licensees may use to assist
them in auditing their measurement control
program.
(10) NUREG/CR·0773, "Training and Qualifying
Personnel for Performing Measurements for theControl and Accounting of Special Nuclear
Material," documents procedures that licensees
may use to assist them in meeting the present requirements contained in the NRC regulations.

RADIOISOTOPES IN I'NDUSTRY

Products Containing Radioactive Material

In fiscal year 1979, the NRC issued two reports onradiation doses associated with consumer products
containing radioactive materials. These reports,

216

NUREG/CR~0215 and NUREG/CR~0216, present
estimates of potential radiation doses to members of
the general public from exposure to timepieces con~taining radioactive material. This radioactive material
is either tritium or promethium~147 mixed with aphosphor used in luminous paint or tritium contained
in sealed~ glass ampules, to be used as a radiolumininescent source. The conclusion that can be drawn fromthe results presented in the reports indicates that theuse of these radioactive. materials in timepieces does
not constitute a significant health hazard.
In November 1978, the NRC issued a report,
NUREG/CP-0001, entitled "Radioactivity in Con~
sumer Products." This report is a compilation of information dealing with radioactivity in products
available in the market place to the general public.
The report is based on papers presented at the
February 1977 symposium on public health aspects ofradioactivity in consumer products.
In July 1979, the NRC issued a supplement to
NUREG-0060, "Generic Environmental Statement on
Routine Use of Plutonium~Powered Cardiac
Pacemakers." The supplement provides updated information on alternative power sources, particularly
lithium batteries, and considers the current extent ofpacemaker use and the makeup of the patient population. The supplement concludes that the results of theoriginal environmental statement are still valid, i.e.,
the routine use of plutonium-powered pacemakers
should be authorized. The supplement further statesthat the plutonium-powered pacemaker provides
physicians with an alternative choice of medical treatment for patients who may require long-term pacing,but who are ill-suited for replacement operations. The
NRC is interested in receiving additional public comments on both documents, NUREG-0060 and its supplement, before considering a final regulation on theroutine use of the plutonium-powered pacemakers.
In 1979, a standard on radiological safety in thedesign and construction of apparatus for gamma
radiography was being developed by a subcommitteeof the American National Standards Institute with
NRC participation. The standard provides guidancefor persons responsible for the design and constructionof apparatus for industrial gamma radiography using
radioactive materials as the energy source. Criteria forthe design of new devices and for qualifying prototypes to performance standards are included. At
year-end, development was not completed, and the
NRC was considering rulemaking proceedings in thisarea.

Licensing Matters
In October 1978, the NRC issued for comment
Guide 6.8, on identification plaques for irretrievable
well-logging sources. This guide describes methods

that would be acceptable to the NRC staff for meetingthe proposed requirements with respect to thecharacteristics and the mounting of permanent identification plaques at the surface of a well that contains
an irretrievable well-logging source. These identification plaques will serve as one aspect of continuing control and are intended to be a long-term indication of asealed radioactive source downwell. Thus, any persons
planning to reenter the well for additional operations
will be alerted to the existence of a source downwell.
The guide was issued for comment concurrently withthe proposed rule changes to 10 CFR Part 30, "Rules ofGeneral Applicability to Domestic Licensing of
Byproduct Material," and 10 CFR Part 70, "Domestic
Licensing of Special Nuclear Material," that will ensure continuing control of byproduct and special
nuclear materials.

NATIONAL STANDARDS PROGRAM
The national standards program is conducted under
the aegis of the American National Standards Institute(ANSI). ANSI acts as a clearinghouse to coordinate thework of standards development in the private sector.
The NRC staff is active in the national standards
program, particularly with respect to setting priorities
so that regulatory views are known regarding thestandards that can be most useful in protecting the
public health and safety. NRC participation is basedon the need for national standards to define acceptableways of implementing the NRC's basic safety regula~tions.
The actual drafting of standards is done by experts,
most of whom are members of the pertinent technicaland professional societies. Approximately 230 NRC
staff members serve on working groups organized bytechnical and professional societies. These societies arelisted in the accompanying table. National standards
are used in the regulatory process only after independent review for suitability by the NRC staff and after
public comments on their intended use have beensolicited and considered.

IAEA REACTOR SAFETY STANDARDSNRC has continued its lead role in organizing andcarrying out U.S. participation in the IAEA program
to develop safety codes of practice and safety guides
for nuclear power plants. The NRC coordinates U.S.
technical activities associated with this program. The
codes and guides will provide a basis for national
regulation by developing countries of the design, construction, and operation of nuclear power plants.
NRC staff members continued to represent the UnitedStates on the IAEA Senior Advisory Group (SAG) that

==================================================217
oversees the program and on the Technical Review
Committees working in the five areas of primary interest: governmental organization, siting, design,operation, and quality assurance. The Director of the
NRC's Office of Standards Development is the U.S.
member of the SAG.
During 1979, the Senior Advisory Group, Technical
Review Committees, and working groups under themdrafted four new guides and completed seven safety

guides that were forwarded to the Director General of
the IAEA with the recommendation that they be
issued. About 40 of the approximately 50 safety guides
planned to date have been drafted and are undergoing
review. During the drafting process, the NRC standards staff coordinated the reviews within the U.S.,
soliciting comments from interested members of the
public, industry, and other government agencies. (Seealso Chapter 9.)

SOCIETIES SPONSORING NUCLEAR STANDARDS DEVELOPMENTACTIVITIES IN WHICH NRC STAFF MEMBERS PARTICIPATE
American Association of Physicists in Medicine
American Concrete InstituteAmerican Conference of Governmental Industrial
Hygienists. American Institute of Chemical Engineers
American Institute of Steel Construction
American Insurance Association
American National Standards InstituteAmerican Nuclear Society
American Society of Civil Engineers
American Society of Mechanical Engineers
American Society for Nondestructive Testing
American Society for Testing and Materials

American Welding Society
Health Physics Society
Institute of Electrical and Electronics Engineers
Institute of Nuclear Materials Management
Instrument Society of America
Metals Properties Council
National Council on Radiation Protection andMeasurements
National Fire Protection Association
National Sanitation Foundation
Society of Naval Architects and Marine Engineers
Welding Research Council

11Regulatory
Research
NRC research continued to address theeffectiveness of radiation measurement
techniques in 1979. Here instruments are
set up In a reactor containment to measureneutron spectra and fluxes.

The lessons for NRC research from the Three Mile
Island accident range from the obvious need to study
the plant itself to the suggestion that a rethinking ofresearch priorities and policies is in order. By the endof fiscal year 1979, several key reactor safety research
programs had been redirected and the Commission
had under study a number of proposals to rechannel or
supplement resources for TMI-related research.
It was clear, for example, that NRC's earlier concentration of research resources on design basis accidents, such as massive loss-of-coolant accidents(LOCA), must be re-evaluated and those resources
either redistributed or augmented to permit greater
emphasis on the smaller kinds of LOCAs and transients and on the operator deficiencies that characterized the TMI event. Better computer codes will beneeded to permit study of the many variations that canoccur in small LOCAs and transients and to predict
with improved precision the behavior of plants in suchsituations. The availability of those codes also will aid
in studies designed to enhance operator capabilities.
Soon after the TMI accident, NRC began to reorient
its 1979 research programs to increase the emphasis in
TMI-related areas, with some $12 million of 1979 program support funds rechanneled into this immediate
research. In addition, fiscal year 1980 research funds
are being reprogrammed, and supplemental funds of$20.9 million are being sought for the TMI-related
effort. With the supplemental funds, the total TMI
research effort in fiscal year 1980 will be about $55.4
million.
Except for the safeguards research program (discussed this year in the separate Chapter 5, "Safeguards"),
all NRC research activities felt the impact of TMI. In
water reactor safety research, the ongoing systems
engineering projects using the LOFT and Semiscale
facilities in Idaho were examined and, where appropriate, deferred to accommodate a new emphasis on
the spectrum of accidents which lies between "designbasis" accidents and "core melt" accidents. The accident at TMI, of course, was in this in-between spectrum.
TMI also pointed up the need for new research on
coolant chemistry-particularly hydrogen evolution-after severe fuel damage, and on the impact ofsevere accidents on plant components, including theequipment qualification and testing techniques used toevaluate them.
In the area of human engineering, new or revised
safety analysis codes now being developed will allow
studies of simulators which could be used in training
plant operators and of instrumentation and control
room display and diagnostic equipment, toward improving operator response to the full spectrum of reactor accidents. In risk assessment research, new event
trees are needed to address accident sequences involving severe core damage and to guide research into theeffect of human error on the course of such sequences.
In parallel with all these studies it is necessary tocontinue investigations of improved plant design
features, including decay heat removal and emergency
core cooling systems, vented containment concepts,
etc., and to intensify ongoing site safety research withnew emphasis on population projections and othersocioeconomic considerations.
These activities are described in the paragraphs
which follow. In addition, details of NRC's implementation of the Congressionally mandated Research to
Improve Reactor Safety are discussed in a required annual report summary at the end of the chapter.

Water Reactor
Safety Research
SYSTEMS ENGINEERING
In early 1979, prior to the Three Mile Island accident, the NRC was beginning to phase down large-

220

break LOCA research to increase the emphasis on
small breaks and other reactor transients. The
Blowdown Heat Transfer (BDHT) program at Oak
Ridge, for example, was restructured to obtainthermal-hydraulic data under slower transient conditions than those previously studied, BDHT instrumentation was modified to cover the slower process, andnatural circulation was studied in a small-scale apparatus. After the TMI accident, many projects were
further modified to include other concerns about small
breaks. Under the Semiscale program, expedited tests
were conducted to simulate TMI-type accident conditions at the height of the crisis, and the facility is now
undergoing major modifications to serve as the major
small-break test facility in the United States.
Numerous other programs are being similarly
redesigned, and these are discussed below.

Integral Systems Tests
NRC's major integral systems test programs arebuilt around the Loss of Fluid Test (LOFT) facility, a
50 MWt experimental pressurized water reactor(PWR) which accomodates both non-nuclear andnuclear tests, and Semiscale, a smaller scale nonnuclear test facility. Both are located at DOE's Idaho
National Engineering Laboratory.

mercial reactors and their emergency core coolingsystems (ECCS).
Significant LOFT events in 1979 included the second large-break LOCA nuclear experiment and a
preliminary small-break LOCA experiment, both performed in May. (The first nuclear experiment was
described on p. 182, 1978 Annual Report.)
The large-break experiment on May 12, 1979, wassimilar to the first nuclear test in that it began with therapid opening of blowdown valves to simulate the instant shearing of the major reactor coolant pipe. The
experiment was conducted at a power level about
1!80th that of a commercial power reactor, yet thecore heat generation rate and coolant pressures andtemperatures were the same as those of a commercial
reactor. The emergency-core-cooling system functioned as expected, although measured fuel cladding
temperatures were lower than predicted. This wascaused by a slug of water passing upwards through thecore during system depressurization, prior to deliveryof emergency coolant.

Loss-of-Fluid-Test (LOFT) Program. LOFT provides experimental information that is used to assess
the analytical models that evaluate the safety of com8[

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NRC's Semiscale facility in Idaho, a small-scale, non-nuclear
model reactor, was called on early in the accident sequence at
TMI to simulate plant conditions as an aid in bringing the plantunder control. This representation of a relief valve leak experimentshows the overlays for pressurizer and vessel liquid levels land forcore cladding temperatures in the time-frame of 5000-7000 seconds(1 hr. 24 min.-l hr. 56 min.) after trip. Oscillations shown forthe pressurizer liquid level were caused by pump cavitation as indicated by the fact that the oscillations subsided when the (»umps
were turned off. Shortly afterwards, the vessel began drainmg andthe vessel liquid level passed below midcore level. When this occurred, the cladding temperatures be~n to rise, indicating void
fonnation even though tile pressurizer liquid level was indicated to
be full.

This overhead view of the LOFT test assembly at the Idaho National Engineerin~ Laboratory (INEL) gives a good indication of
the size of the facllity. The apparatus contains a 50 MW(t).
volumetrically scaled PWR system which permits NRC researchers
to study the response of the engineered safety systems to loss-ofcoolant accidents. Fiscal year 1979 saw a considerable reorientation of LOFT experiments toward the study of small-break accident phenomena such as those prevalent at Three Mile Island.

=============================================221
The Semiscale Mod-3 features a 3.658-meter core, upper-head injection capabilities, active broken loop components, a full-length
V-tube broken loop steam generator, thm-core densitometers andother advanced instrumentation with more than 700 potential
measurement locations. One type of instrumentation IS the intact
loop spoolpiece, shown in cross section. The spoolpiece consists of abidirectional turbine, a resistance bulb thermometer a watercooled pressure sense line, a fluid thermocouple and 'metal thermocouple, a two-beam cesium source densitometer and either asin21e water-cooled transducer drag disc or a three:pin watercooled transducer drag screen. It can be installed at various points
along the intact loop, as shown.

Type 2steam
generator---....

Chromel·alumel
thermocouple probe
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gamma densitometer
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INEL·5-6489-1

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downcomer

The first two large-break nuclear experiments yielded results that were quite similar. What differences
there were stemmed from performing the second test
at a core power 50 percent greater than that used forthe first test, since this increased the coolant
temperature rise in the core some 45 percent.
Temperature response of the fuel in the two largebreak nuclear experiments, and the post-test core requalification, indicate there was no fuel damage ineither test, hence, that fuel is available for futuresmall-break tests. As noted earlier in this chapter,
small-break LOFT experiments were planned to help
answer questions related to the small-break accident atThree Mile Island.
The first such experiment, conducted on May 31
with the nuclear core at zero power, yielded data
which were needed to prepare for the first nuclear
small-break test in November, 1979. It consisted of aslow depressurization during which the ECCS performed as expected, and it provided new informationon accumulator injection and natural circulation. The
staff expects the small-break and anomalous transient
tests, using more sophisticated control-room diagnostic instruments, to reveal new information about
the response of a reactor facility to accidents similar tothe one at Three Mile Island. The information should
also suggest what operator actions are necessary to
mitigate the consequences of such accidents.

Foreign participation in the LOFT program continued with representatives from Austria the Federal
Republic of Germany, Japan, the Netherlands,
Switzerland, and Scandinavia actively participating in
the program.
Semiscale Program. The Semiscale facility nuclear
core is simulated by electrically heated rods whosegeometry corresponds to the nuclear fuel rods in a
typical PWR. Using the Mod-3 configuration (see p.
150, 1977 NRC Annual Report), the Semiscale facility
was modified quickly during the recovery from the accident at Three Mile Island and used to simulate conditions in the plant in evaluating alternatives for securing the reactor. During critical phases of the TMI accident, Semiscale tests provided significant insights into
the growth and movement of noncondensible bubbles
such as those caused by hydrogen. In post-accident
analyses, Semiscale test are being used to model the sequence of events in continuing investigations of TMItype behavior. In all, ten tests were done to furnish additional information on the Three Mile accident.
Other highlights of the Semiscale Program during
1979:• Completed small break tests to assess licensing
codes used to analyze a postulated accident. (Thisis the test series which was expanded to include
the Three Mile Island accident phenomena).

222• Completed baseline tests on the Mod 3 facility toestablish performance parameters of the system in
evaluating different phases of a LOCA. Nine
published NRC reports summarized these tests.
• Tested the capability of a prototype optical probe
to characterize water mixing behavior in thevessel downcomer. It was successful.
• Completed and published the results of studies
scaling Semiscale MOD-3 to PWR, LOFT to
PWR and Semiscale MOD-1 to LOFT. Results
confirmed the rationale earlier employed in scaling LOFT from PWR and Semiscale from LOFT.
• Published the report on Semiscale tests andanalyses performed in support of the Three Mile
Island evaluation effort.

Separate-Effects Experiments
Major test facilities for NRC's separate effects experiments are the Thermal Hydraulic (Blowdown)
Test Facility (THTF) at Oak Ridge National
Laboratory, Tennessee; FLECHT SEASET* at
Westinghouse in Pittsburgh; the Two-Loop-TestApparatus (TL T A) and Sector Steam Test Facility(SSTF) at General Electric in California, and thesteam-water mixing facilities at Battelle Memorial Institute in Ohio. Bench-scale tests and instrument
development programs also are being conducted atseveral university laboratories. A table summarizing
the location and capabilities of these facilities is on
p. 151, 1977 NRC Annual Report.
Two Loop Test Apparatus (TLTA). TL T A tests during 1979 showed that the injection of emergency corecooling water cools the bundle during the early phasesof the postulated accident more effectively thanpreviously believed. Studies of later phases of the
LOCA and tests of other, more probable accident sequences are planned for 1980 and beyond.
Steam Water Mixing Tests. The studies of steamwater mixing effects on the penetration of coolingwater in models of PWR vessels conducted over the
past five years in the small 1115 and 2/15 scale models
at Battelle Columbus Laboratories and at Creare, Inc.
were largely completed during fiscal year 1979. A major result has been the accumulation of additional
evidence of the conservatism in models used in thelicensing process. Knowledge gained from these smallscale tests now will be used in the planning andanalysis of full-scale penetration tests to be conductedin the Upper Plenum Test Facility in Germany. (See"3D Program," under "Research Support," later inthis chapter.)

Full Length Emergency Cooling Heat Transfer Separate Ef.
fects/System Effects Tests.
oj<Counter Current Flow Limit (CCFL) Refill/Reflood
Program. This project, sponsored jointly by NRC, theElectric Power Research Institute (EPRI), andGeneral Electric, was initiated in 1979 and features
the Sector Steam Test Facility (SSTF) , a full-scalemodel of a 30 degree sector of a boiling water reactor(BWR). Tasks undertaken to date include investigations of the distribution of cooling water sprayed over
the top of a core and how that cooling water
penetrates fuel bundles. A modeling effort is in process
to aid the development of a BWR version of the TRAC
code (see "Code Development," below).
FLECHT-SEASET Program. This program is
described in detail on p. 184, 1978 NRC Annual
Report. In 1979, Westinghouse completed separate
effects studies of both steam generator and core
behavior during reflood, and continued the studies ofcore blockage using a small 21-rod bundle. The latter
program will be expanded to use a 161-rod facility toinvestigate bypass as well as blockage geometry, andto permit comparisons of results with core blockage
data from the Cylindrical Core Test Facility in Japan(see "3D Program," below). These tests should clarify
some uncertainties regarding the conservatism of heattransfer-rate and flow-regime criteria now used inlicensing regulations. The system components (corerod bundles and steam generators) which have beenstudied separately will be integrated to investigate
system interactions during various post-accident cooling tests. These system tests will focus on heat transfer
mechanisms that are important in both large andsmall-break LOCAs. Steam heat transfer data fromthe 161-rod facility were used in TMI-related activities.
PWR Blowdown Heat Transfer Program (BDHT).
The PWR Blowdown Heat Transfer Program at Oak
Ridge was redirected in 1979 toward gaining a better
understanding of fluid conditions in a PWR core during slower depressurization such as that characteristicof TMI. After modifications, ORNL ran a series oftests with the electrically heated 7x7 fuel-rod bundle(Bundle No.2) to simulate various transient conditions. Initial tests using Bundle No. 3 were run in
December 1979. This bundle features much-improved
instrumentation and should produce a considerably
improved understanding of pressure, flow, power, andtemperature behavior. Tests with Bundle No.1 were
described on p. 183, 1978 Annual Report.
Model Development Studies. NRC continues to use
small scale tests to study the various phenomena
associated with steam-water flow. Each test is
directed toward a particular effect to produce improved models or better data for input to code calculations.
In 1979, Argonne National Laboratory completed experiments on the effect of controlled oscillations on
reflood heat transfer; Massachusetts Institute of

223

Technology completed work on spacer-grid effects on
reflood and on natural circulation flows; Northwestern University continued its experiments on therate of steam condensation, and Brookhaven National
Laboratory continued its study of vapor generation
rates in depressurization situations. In addition,
Rensselaer Polytechnic Institute and Argonne National
Laboratory undertook the development of models
dealing with void fractions in reactor coolant flow
during accidents, and the State University of New
York at Stony Brook began work on models for theamount and character of droplet flow in and above areactor core.

NRC's PWD Blowdown Heat Transfer Program at Oak Ridge,
Tennessee, entered a new phase in 1979 as tests with fuel-rod
simulator Bundles 1 and 2 were completed. Shown above is Bundle
1 after removal from the Thermal Hydraulic Test Facility(THTF). Below: Bundle No.3, an 8x8 rod array, is lifted into avertical position prior to insWlation in the THTF. Bundle 3 hasmore than 1200 instrument sites which can be monitored 20 times
per second during experiments.

Instrument Development. Test facilities require
many sophisticated instruments which are not commercially available. NRC's advanced instrumentation
program, described in the 1977 and 1978 Annual
Reports, is designed to fill this void. Progress during
1979 was reported in July at the Reactor Safety Instrumentation
ReviewGroup
Meeting(NUREG/CP.0007). A summary of that reportfollows:
A technique called "pulsed neutron activation" wasdeveloped by Argonne National Laboratory tomeasure the velocity and density of steam/water flow.
The technique will be used as a standard calibrationfor other instruments. Sandia Laboratories developed
a portable neutron generator as part of the U.S. contribution to the international "3D" program (seebelow). Flow-measuring instruments, including turbine, drag disk and gamma densitometers have beendeveloped and improved by INEL, and impedance

MULTICHANNELANALYZER

A July 1979 progress report on the NRC advanced instrwnentation
program described the pulsed neutron activation (PNA) techniquefor measuring velocity and density of steam-water flow. The
technique involves: activating fluid with a neutron pulse from a
portable neutron source (left); detecting activated nuclei a known
distance downstream, and plotting the transit-time; determining
average mass flow velocity and the average density of the fluid,and deriving mass flow from the latter two averages. The torus
detector (right) features sodium iodide scintillators mounted tophotomultiplier tubes. The plot at center is reproduced from PNA
measurement of an air-water mixture. PNA equipment has been
installed at the Semiscale and the LOFT facilities, and will be
used in tests with other devices in 1980.

224

probes have been developed by ORNL. Other advanced instrumentation for flow measurements include optical probes, ultrasonic densitometers, and a stagnation probe, developed by INEL for measurements in
LOFT. Rensselaer Polytechnic Institute, State University of New York at Stony Brook, Lehigh University,
and Northwestern University continued their development of void fraction probes, film probes, special
thermocouples, laser doppler techniques, and pitot
tubes to better measure the thermal hydraulic properties of water and steam under accident conditions. So
far, all of this advanced instrumentation technology
has been applied in research facilities. Evaluation of
its applicability to commercial power plants continued
in 1979 and into 1980, particularly in light of an important lesson learned from the Three Mile Island accident-the need for instrumentation to measurewater levels in reactor cores.

FUEL BEHAVIOR
NRC's fuel behavior research programs in 1979 included cladding experiments, in-reactor tests, fuelmeltdown and fission product transport tests, and fuelcode development. These are described below. Data
produced in these programs in previous years contributed significantly to improved understanding andanalyses of the TMI-2 accident, particularly in theareas of oxidation, embrittlement and ballooning of
the cladding and the release of fission gases from thefuel. As technical data from post-TMI analyses
became available, the research staff began to recast
some of those programs toward the study of fuelbehavior when fuel rods are uncovered or severely
damaged. In some tests, fuel assemblies will be allowed to boil dry to ascertain whether or how best they
can be cooled. In others, the release and transport offission products from damaged fuel will be studied. 'Some of the facilities and techniques which will be
used for these TMI-related projects are described in
the following summary of ongoing fuel-behavior
research activities.

Cladding Experiments
Multirod Burst Test (MRBT) Program. The MRBT
program at ORNL has two main objectives. The first is
to better define the deformation behavior of unirradiated Zircaloy cladding under conditions
postulated for a loss-of-coolant accident (LOCA). The
second is to provide a data base for use in assessing
geometrical changes which occur in fuel rod cladding
and the extent of coolant flow restrictions that thechanges (such as ballooning and rupture) might produce.
Data accumulated in the MRBT program havecome from unheated-shroud single-rod tests and from.heated-and unheated-shroud multirod (4 x 4 bundle)

Tube burst test equipment at the Battelle Columbus Laboratory in
Ohio is used to determine changes in fuel rod cladding burst pro-perties resultin~ from irradiation in a commercial power reactor.
Above, techniCian (seated) operates the tube burst console which
prosrams, reads and records test parameters, while a technician(right) uses remote manipulators to load a radioactive specimen
into the test stand in a hot cell. Below, a specimen is mounted in
one of two test fixtures. A furnace for conducting elevatedtemperature burst tests is shown above the test specimen.

225

tests in which the shroud temperature at the time of
burst was no closer than 80°C from the fuel bundle
temperature, and in which the rods were not restrictedin their outward movement. The importance of closely
simulating these thermal and confinement factors is
not clear at this time, but the data base will be increased by investigating these concerns in tests. One conclusion that has held up throughout the testing is that, at
a given temperature, the temperature gradients determine the extent of deformation, i.e., the more uniform
the temperature distribution, the greater and more
uniform the deformation.
Mechanical Properties of Zircaloy. Work in 1979 at
Argonne National Laboratory produced information
which will lead to better criteria for predicting
damage to cladding embrittled by oxidation at hightemperatures. These new criteria are stated in terms of
measurable mechanical properties which can be used
to define the oxidation limits that will permit the cladding to survive such conditions as thermal shock, impact and physical damage, and the loads on cladding
resulting from the interaction of reflooding water andsteam.

In-Reactor TestingPower Burst Facility (PBF). At the PBF in Idaho,
research continued in 1979 on the behavior of fuel rods
under various operating and accident conditions. (Seepage 154, 1977 Annual Report.) The "reactivity initiated accident" (RIA), described on p. 188, 1978 Annual Report involves a burst of power generated whena control rod is ejected from the core. Early in fiscal
year 1979, two RIA experiments were performed in
the PBF test reactor. In the first test, two fresh fuelrods and two pre-irradiated fuel rods were exposed toa power burst that in earlier experiments had done

This LOFT Lead Rod Test Assembly fabricated by EG&G Idaho,Inc. for testing in the PBF, contains lour individually shrouded
and instrumented PWR-type fuel rods. The apparatus is used to

substantial structural damage to the test fuel. In someregions of the pre-irradiated rods the fuel swelled morethan expected, and the swollen fuel and debris fromadjacent parts of a rod blocked the flow shroud morethan expected.
The second PBF RIA test was performed with four
pre-irradiated rods previously exposed to a lower
power burst near the level at which cladding could be
expected to fail and to release fission products. Only
one of the four rods failed, but a series of small
longitudinal cladding cracks appeared which resembled the kind of pellet-cladding interaction whichusually induces cladding failure. Since this rod had notbeen opened after pre-irradiation, and the companion
rods had been, these effects will be studied in future
RIA damage threshold tests using unprocessed preirradiated rods.
Later in the year, three LOCA blowdown tests were
performed to confirm that planned LOFT tests with
unpressurized rods could be run without jeopardizing
the fuel bundles.
The final two PBF tests of 1979 used both fresh andpre-irradiated rods, pressurized to match the fill gas
pressures of commercial rods when new and at or near
the end of life, to examine the "ballooning" associatedwith cladding rupture with the peak temperatures in
the 1050 0K-15000K (1430°F-2240°F) range. In one
test, circumferential ballooning as high as 50 percent
was observed. (NOTE: More than 70 percent ballooning of four adjacent rods is required to block a flow
channel in a commercial power reactor, and somecooling water flow through any cladding ruptures ispossible even then.) The second test was performed in
the last months of fiscal year 1979, and the rods had
not been examined at year's end.

collect data for the assessment of predicted fuel behavior duringloss-of-coolant accident testing in LOFT.

226====~==========================================

Fuel Meltdown/Fission ProductRelease and Transport
Fission Product Release and Transport research continued at two laboratories:• Oak Ridge National Laboratory completed an experimental program to determine the quantity
and type of fission products which might escapeduring reactor accidents. In four hightemperature tests, segments of irradiated commercial fuel rods were subjected to peaktemperatures (1200° to 1600°C) well above thosepredicted for successful control of a LOCA. These
tests, conducted in a flowing-steam environment,
indicated that the release of fission products
iodine and cesium increased tenfold and of krypton by a factor of five within the temperature
range 1350°C to 1400°C. The results of tliese tests
were helpful in estimating the fuel temperaturesattained during the TMI-2 accident.
• Battelle Columbus Laboratories published auser's manual for the TRAP-MELT code, whichmodels the transport behavior of fission products
in primary coolant systems of LWRs in various
accident conditions, including fuel meltdown.
Core-Melt Research. Sandia Laboratories continued
investigating the potential for thermal explosions, if

A combined experimental Ianal~c program at Argonne National
Laboratory, aimed at better understanding the effects of fission gas
releases on irradiated fuel, employs this direct electric heating apparatus (left) and predictive models of the GRASS family of computer codes. In the experimental effort, fuel pellets are heated totemperatures similar to nuclear heating, then studied in various
ways, including the use of scanning electron microsco~ photography. The photograph reveals an interconnected network of tunnels
on the grain boundaries of irradiated UOa fuel.

molten core materials were to contact water, with aseries of experiments dealing with the efficiency of
converting the thermal energy of the melt intomechanical energy. In a series of 48 tests using molten
iron/alumina with masses up to 27Kg, the maximum
efficiency was measured at 1.34 percent, a factor of
about 20 less than the maximum theoretical efficiency
for thermal interactions.
Another Sandia Laboratories investigation exploredthe interaction of molten core materials and concrete,
producing important data on the gases and aerosols
generated, the penetration rate of the melt into theconcrete, and the rate of fission product evolution
from the melt. This information was used to develop
an advanced melt/concrete computer code called
CORCON.

Fuel Behavior Codes
Fuel Rod Analysis Program (FRAP). Informationfrom the PBF, LOFT (see above) and Halden Reactor
Project (see p. 189, 1978 Annual Report) programs isused in developing and assessing NRC codes "FRAPCON ," used for the steady-state analysis of fuel rod
response during normal reactor operation, and FRAP-

===================================================227
T, used for transient analysis of fuel rod response during off-normal reactor conditions. During 1979, improved models and correlations gave FRAP-T acapability to more accurately predict fuel rod
temperatures, deformations, and possible failures during all phases of a LOCA. A report on the revised
library of materials properties (MATPRO-ll) neededby the two codes was issued (NUREG/CR-0497), andat the end of the reporting period, both FRAPCON-land FRAP-T5 were available at the National Energy
Software Center for distribution.

small-scale research tests to the behavior of full-scale
power reactors. In 1979, the initial version of the
TRAC code (applicable to pressurized water reactor)
was sent to the National Energy Software Center.
Work to improve that version and work on the boilingwater reactor version continued into 1980. In addition, the TRAC system code is being linked with moredetailed component codes such as fuel element andfuel bundle codes.
Development of the WRAP code (see below) package used in licensing audit calculations for LWR
plants was completed.

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RELAP-4. MOD 7 of this system's LOCA code has
been applied extensively in analyzing the TMI "accident, as well as in performing TMI sensitivity studies,and in auditing small-break calculations developed byvarious steam system suppliers. The primary conclusions from this analysis were as follows:• The RELAP-4/MOD 7 computer code addresses
most of the phenomena thought to occur in afull-scale pressurized water reactor under TMItype conditions.
• A detailed input model was needed to calculate
the phenomena occurring during the TMI accident, and this resulted in large computer calculational times.

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Measured fuel center temperature (K)The development of methods for the presentation of data andcalculated results is an important aspect of the fuel assessment pro..
cess because of the vast amounts of mformation that must he interpreted. This graphic presentation illustrates one way in whichassessment results may he presented. It shows a predicted vs.
measured plot of fuel temperatures obtained using the FRAP computer code. Such plots show the code assessor and user not only thedeviations from perfect agreement (the 45 0 line) but also trends inthe data (horizontal axis) and prediction (vertical axis) spreads, including areas where data are lacking, more abundant, or changed.

COMPUTER CODE DEVELOPMENT
Code Development
Research results from the reactor safety research
program are used to develop computer codes andassess their accuracy in analyzing nuclear plant accident. Thus, code development is a focal point for reactor safety research results.
The reference computer code (RELAP) for analysis. of reactor system accident behavior is in worldwide
use. Development and improvement work on RELAP
is being phased out, and the major code effort now isaddressed to the advanced system code, TRAC. TRAC
contains better and more complete descriptions ofphysical processes, which makes the code less susceptible to scaling errors in extrapolating from results of

WRAP. The Water Reactor Analysis Package(WRAP) is a set of evaluation-model computer codes
linked together to perform accident calculations in thelicensing review process. WRAP-BWR model wastested in 1979 in a calculation of a typical BWR plantand on test data from the Two Loop Test Apparatus. A
computational module for the refill portion of a
postulated LOCA was added to WRAP-PWR model,
and the complete PWR package was exercised in a
calculation of a typical PWR plant, and on test datafrom the LOFT facility.
TRAC. The Transient Reactor Analysis Code(TRAC), an advanced best-estimate computer program, is designed to predict the thermal and hydraulicresponses of LWRs to LOCAs and other transients.
(For a detailed description of TRAC, see Page 159 of
the 1977 Annual Report).
TRAC was used extensively in analyzing the Three
Mile Island Accident. The scenario of events and accident parameters-to the extent they were
known-were modeled, and these factors were thenvaried in additional calculations to learn what would
have happened had different decisions been made during the accident. The TRAC computations indicatedthat the plant safety systems themselves were adequate
to prevent the core from being uncovered and causing
core damage, and that if the high pressure coolant injection system had been left on, the core would havestayed covered with water. Other activity involving

228---Data

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models for analyzing fuel rods, jet pumps, and steamlines. In a companion move, NRC discontinued its
development of the advanced system code "THOR"(See p. 193, 1978 Annual Report) following a review of
three candidates (THOR, RELAP-5, and simplified
TRAC) for a fast running systems code.

Code Assessment

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Time
One of the objectives of code assessment is to detennine the accuracy of a code prediction (or the capability of the code to makea prediction). The procedure illwtrated here, which is Wlder
development at EG&G Idaho, Inc., shows experimental data intenns of mean value and Wlcertainty boWlds. The calculation, asinsde-value curve, is compared with the data range. The accuracyof the calculation is detennined by integrating (summing) the
deviations of the calculation from the mean and nonnalizing(dividing) the total error by the width of the Wlcertainty band of
the data. Shaded areas illwtrate ways in which the calculation
may deviate from the data.

TRAC included publication of Volume II of the
TRAC-PIA manual. It details test data analyzed by
the code during its development.
Work also was started on a much faster version of
TRAC, TRAC-PFl, scheduled for release in 1980. Thisversion can reduce the number of computational nodes
required for a given calculation, thus speeding therunning time. Also, a revised numerical method ofsolving differential equations will accommodate larger
time-steps during slower transients similar to the onethat occurred at TMI.
TRAC calculations were performed both in pretest
predictions and post-test analyses on tests performed in
the LOFT, SEMISCALE, and other experimental
U.S. facilities, and in support of German and Japanesefacilities as part ~f the International 2D/3D
Refill/ Reflood Program.
COBRA. The COBRA-TF multidimensional vessel
code (described on p. 191 of the 1978 Annual Report)
has been linked as a vessel module to the TRAC code toobtain self consistent vessel boundary conditions during the course of a postulated accident. The linked
code has been named COBRA-TRAC.
RAMONA-III. This code, developed by SCANDPOWER, Kjeller, Norway, is being employed at
Brookhaven National Laboratory (BNL) in calculating
BWR operational transients. The code accommodates
up to 100 parallel thermohydraulic channels with different flows, voids, and conditions. While working
with RAMONA-III, BNL also developed improved

A number of codes were assessed during the fiscal
year. An assessment of RELAP-4/MOD 6 was completed at INEL with the publication of an assessment
report in December 1978, and an addendum report in
September 1979. An assessment procedure and user
guidelines were developed, and multiple comparisons
between calculations and data were obtained for the
Semiscale MODs 1 and 3, LOFT, PKL, Marviken,
PBF, THTF, FLECHT and FLECHT-SEASET facilities. In systems comparisons, clad temperature
calculations tended to be too high when compared toexperimental data, indicating need for improvements
in calculations of certain heat transfer and fluid flowpheonmena. These are being addressed in the RELAP4/MOD 7 code version.
A coordinated assessment program for the TRACPIA code was initiated at LASL, BNL and INEL, inwhich particular emphasis is placed on "blind" predictions-predictions in which the result of the experiment is not known to the analyst at the time of
preparation of the code input. As an example, all
LOFT nuclear test applications in the assessment of
RELAP-4/MOD 6 and TRAC·PIA have been doneusing blind predictions.
The assessment of the TRAC code is a long-term
project and a plan has been prepared based on three
types of tests: Integral Systems Tests (for example
LOFT and Semiscale), Separate Effects Tests(separate components of the reactor system), and Basic
Tests (designed to explore basic thermohydraulicphenomena that are of importance during a LOCA).
Specific experiments have been selected and many ofthese were run during 1979. The assessment program
will continue in 1980.

METALLURGY AND MATERIALS
NRC's Metallurgy and Materials Research Programis designed to provide confirming information regarding the integrity of water reactor vessels, tubes andpiping syste~s. It encompasses research in fracture
mechanics including design criteria for fracture
prevention, flaw detection and evaluation, the growth
and arrest of cracks; radiation embrittlement; corrosion, and the uses of acoustic, electro-chemical andother techniques to help assure component integrity.
Large vessel tests in 1979 continued to validate thedesign conservatisms in reactor pressure vessels. In addition, fracture mechanics research demonstrated theintegrity of weld repair techniques and identified con-

===================================================229
ditions under which thermal shock will not propagate
cracks.
The impact of Three Mile Island on these programs
will be known in detail once cleanup plans for theplant are developed. Steps already are being taken,
however, to integrate NRC research requirementswith those of DOE and other agencies in post-accident
examinations of the containment building and safetysystems and components.

Fracture Mechanics
Elastic-plastic Fracture Mechanics. Reactor vessel
steels are designed to be extremely tough in reactor
thermal environments so that sudden, rapid, brittle
fracture will not occur. If there is a crack in the vessel
wall, however, possibly caused by corrosion, there isa
chance that the crack will grow through a process called ductile tearing. Such tearing is "stable," that is, itwill continue slowly if the internal pressure is highenough to sustain it. The tearing ceases at lower
pressures. It is important to study the nature of these
phenomena since conditions may develop wherepressure loads are high enough and cracks havebecome large enough to make the crack growth"unstable," i.e., the ductile tearing propagates at ahigh rate and endangers the integrity of the vessel orpipe.
The "tearing-instability" method used for structural
analysis, which was developed at Washington University in St. Louis, has been found useful in determining
whether pipe and pressure vessels cracks will grow in astable manner when subjected to severe loads such as
those generated by earthquakes, water hammer effects, etc. It also will permit characterization of theresistance to crack growth of irradiated materials.
These phenomena are discussed below. A parallel effort at the Naval Ship Research and Development
Center in Maryland is the development of a newtechnique for measuring "J-resistance" curves (fracture toughness resistance to ductile tearing for piping
and vessel steels), using the tearing-instability concept.
Fracture Toughness of Irradiated Materials. During
the life of a power-reactor pressure vessel, it is bombarded by high-energy neutrons, which reduces thefracture toughness of the vessel steel. An NRC program at Oak Ridge attempts to quantify loss of
toughness to establish margins of safety against fracture. This has involved the irradiation of several hundred specimens of reactor steels and weld materials,
ranging in thickness from one-half inch to four inches.
Approximately 50 specimens of weld material havebeen tested, and it is anticipated that the results will
influence licensing decisions and safety evaluations ofcommercial reactors, as well as contribute to theresolution of the generic issue addressed in Task A-IIof the-unresolved safety issues. (See Chapter 3.)
Thermal Shock. The fifth in a series of experiments
at Oak ~dge was run in 1979 to demonstrate the

Researchers at Oak Ridge have perfonned simulated weld repairson thick-walled cylinders and pressure vessels as part of NRC-s
Heavy-Section Steel Technology Program to assess a method
recommended in American Society of Mechanical Engineers(ASME) codes. The method is designed for in-service repair ofcomponents when it is not practical to apply the high-temperaturestress-relieving process usually employed. The pressure vessel
shown here was pressurized to failure with a large crack
deliberately implanted where crack propagation was most likely
under combined repair and pressure-loading stresses. The test
demonstrated the importance of residual stresses to vessel integrity,
but it also showed that failure under complex loading conditions
can be predicted, a finding which improves the system of techniques available for evaluating pressure vessel safety. The tests
have also demonstrated that the structure itself, under some circumstances, will arrest rapid propagation of a crack.

structural integrity of a reactor pressure vessel whensubjected to the emergency injection of cold water
during a postulated accident. This experiment indicated that an existing crack was arrested, without
penetration of the outer cylinder surface. Further tests
with large diameter cylinders to more closely approximate the performance of full-sized reactor vessels will
be conducted in 1980.
Reliability of Piping Systems. The continuing incidence of cracks in piping has demonstrated the needfor research to quantify and upgrade pipe-system
reliability. Programs initiated during the year include
development of toughness and crack growth rate dataon typical piping; and estimates of the size, distribu-

230tion and detection efficiency of flaws and reevaluationof stresses, loads, and potential pipe-break locations.
At year's end, data were being combined for a probabilistic analysis of NRC pipe design and operation
criteria and for reconfirming the reliability of nuclear
piping systems.

Steam Generator Tube Integrity
Stress Corrosion Cracking of PWR Steam Generator
Tubing. The NRC research program initiated at
Brookhaven National Laboratory in 1978 to develop adata base for predicting stress corrosion cracking insteam generator tubing was supplemented in 1979,
using tubing that cracks abnormally quickly in addition to the regular tubing used in actual steam
generators. This supplemental testing will permit
more rapid validation of the predictive models now
being developed. Also in fiscal year 1979, otherbaseline data were under development which would
allow predictions of cracking as a time-function oftemperature in high purity water. Tests of various
water chemistry effects also were started.
Integrity of Flaw~d Tubing. Battelle Pacific Norfh·
west Laboratory's program to investigate
burst/collapse strength of flawed steam generator tubing contributes to Tasks A-3, 4 and 5 of the unresolved
safety issues. (See Chapter 3.) The first of three major
phases was completed in 1979 with results showingthat even grossly defected tubing retains considerable
strength. Although some eddy current tests showedthat the techniques used may not accurately
characterize the tubing defects, the overall conclusions
were that pertinent NRC criteria are appropriately
conservative. Research thus far has used mechanicallydefected tubes, but plans were being made at the endof 1979 to use service defected tubes.
Research at ORNL to improve eddy current test
methods, and instrumentation for use during inservice inspections of tubing has already produced
computer codes that can optimize examination instrument designs. Prototypes of these designs were being
evaluated on machine-flawed samples at the end of theperiod.

Radiation Embrittlement
Irradiation Anneal-Reirradiation Program. Research on periodic heat treatment (annealing) of irradiated pressure vessel steel has been reported in NRC
Annual Reports for several years (see p. 196, 1978
report). Work performed this year confirmed that
periodic, one-week heat treatments of irradiated steel
at 200 OF above the normal operating temperature will
give continuing relief from radiation embrittlement,
but the same treatment at only 100°F above normal
gives only temporary relief. The implications of heat
treatment at this temperature were under study at theclose of 1979.
6

Irradiation Surveillance Dosimetry. To assess thesignificance. of radiation embrittlement caused byneutrons emitted by reactor fuel during operation one
has to know the numbers and damaging effects of theneutrons. At Hanford Engineering Development
Laboratory in Washington, with major contributions
from Oak Ridge National Laboratory and the National
Bureau of Standards, a benchmark test facility
simulating a pressure-vessel wall was completed in1979 and experiments dealing with the numbers ofenergy levels of neutrons at specific locations wereconducted. The results showed very good agreement
between measurements and calculations, and this willaid in future estimates of reactor vessel Hfetimes, andin the preparation of standards for calculating andpredicting neutron flux and spectrum in operating
power reactors.

Flaw Detection
Acoustic Monitoring. Continuous on-line monitoring of reactor components during operation using theacoustic emission technique has been under study atBattelle Pacific Northwest Laboratories (PNL) .
Acoustic emission "hears" the waves produced by a
growing crack in a reactor component, and by using
several sensors distributed over the component, thecrack can be located. In 1979 capabilities also weredeveloped to evaluate the severity of a flaw and to
distinguish different flaw types from acoustic emission
signals. In 1980, these developments will be examinedin a large structural test that will simulate all important reactor operations. Later a commercial reactor
will be instrumented for final validation and acceptance of the techniques.
The CARD, Inc., program to develop acoustic emission techniques for the detection and location of flaws
produced during welding (p. 197, 1978 Annual
Report) was completed in 1979. The resulting fieldequipment will detect and locate flaws during
welding, as well as distinguish among different typesof flaws and evaluate their severity.
Vibratory Excitation. Another continuous monitoring technique under study utilizes internal friction of amaterial to detect certain kinds of cracking in
stainless-steel reactor pipes. This technique, developedby Daedalean Associates of Woodbine, Md., depends
on identification of microscopic changes in thematerial as the precursors to cracking.
RES~ARCH SUPPORT

Research Support in 1979 encompassed Operational
Safety Research work, and Program Management
Support and Technical Support activities. Operationalsafety research includes fire protection, qualification
testing evaluation, noise diagnostiCS, human factors,
and assessment of the behavior of safety. and relief

231valves under certain postulated accident conditions(now particularly oriented to a TMI-type of accident).
Program Management Support is provided for certain
areas of international research. Technical Support includes the dissemination of safety research informationand computer codes.

Operational Safety Research
Fire Protection. NRC's fire protection research wassubstantially reoriented in 1979 with project researchpersonnel shifting to work on a series of full-scale firetests. These tests will mock up portions of actual plantfacilities (such as a cable spreading room) and subject
them to design basis fires to test detection and suppression systems. The first full-scale test will take place in1980 .
.In a fire confinement study, six tests dealing withthe relationship between ceiling/wall materials and afire in a tray of electrical cables showed that the closer
the tray is to ceiling/wall corner, the hotter the cable
will be, and the greater its weight loss. Also there ismore chance of fire propagation between cable trays.
The new fire suppression test facility at Sandia
Laboratories in New Mexico was finished in 1979, andcheckout tests have been scheduled. The facility will
be used to study the effectiveness of fire suppression
agents (Halon, water and carbon dioxide).
Qualification Testing Evaluation. In 1979 Sandia
Laboratories completed a new facility to test
qualification methodologies. The facility features irradiation, steam, chemical, pressure and temperaturecapabilities. A series of facility checkout tests as well as
the first test of electrical cable connectors from a commercial nuclear power plant already have been run.
Plans are under way to test additional safety-related
equipment under LOCA and MSLB conditions. The
adequacy of the test methodology will be evaluated inthese tests. In another study completed under this program, final calculations were made on the capabilityof a cobalt-60 radiation source to simulate LOCA accident radiation effects.
Other research activities included the continuing
assessment of methods used to predict the behavior ofelectrical cables when "aged" by exposure to heat,humidity and radiation. A model has been developedthat can be used to predict useful material life by accelerated thermal and radiation aging, including thesynergistic effects of their combined application. Plans
have been made under this program for participationin the examination of safety-related electrical equipment from TMI-2.
Noise Diagnostics. Noise diagnostics research aims
to improve the monitoring of the behavior of a nuclear
power plant by measuring the noise associated withsuch signals as vibrations, power oscillations, loose

Government testing in high teclmologies may be conducted at very
elaborate and expensive facilities, such as the FFTF, or on test.
beds designed for specific short-term experiments such as the ap}Jaratus shown here. The llhoto, taken during the conduct of an
EG&G Idaho, Inc., test, Shows the government-owned Transient
Test System (TTS) located at Wyle Laboratories, Norco, Cal., afacility designed to test flow-measurement instrumentation undertransient two-phase flow conditioru identical to those of the LOFT
system.

parts, etc. In 1978, researchers at Oak Ridge completed a code that statistically combines noise sources
and that has done a reasonably good job of predicting
power/noise spectra (signal as a function of frequency)
in a nuclear power plant. It is known that anomalies
such as instrument tube vibration or core barrel
motion changes the spectrum of reactor noise. Once anoise change is understood, one can use computermodels to ascertain the abnormality causing it.
Human Factors. In 1979, researchers at Oak Ridge
completed the first phase of a study of operator
response to accidents. Study of events at several plants
show that adjustable time margins may provide themost reasonable criteria for operator response, since
experienced operators tend to overestimate their ability to respond quickly. This work now includes various
post-TMI studies of the use of control-room simulatorsin training reactor operators. A report including
recommendations for the future use of simulators will
be published in 1980.
Safety/Relief Valve Behavior. A literature study todetermine the flow behavior of safety and relief valves
under ATWS conditions (See Chapter 3) was published in March 1979 by the Energy Technology

232

VlNT TUBE
THEIIMOCOUPLES

In-Core Guide Tube
Impedance Measuremm Assembly
Flag Probe Type
~O(2t;/ 7U

NRC programs at the Idaho and Oak Ridge National Laboratories
include die development of instrument systems for international(3D) activities in Germany and Japan. Some instruments andtesting activities under these programs are: (1) An in-core guidetube impedance measurement assembly developed at Oak Ridgefor the PKL facility in West Germany. (2) Testing the assembly ona steam-water test stand at Oak Ridge. (3) Instrumented spool.
piece, featuring 11 transducers, developed by EG&G in Idaho for

the PKL. Four such devices have been sent to Germany and eiKht
to Japan for the Cylindrical Core Test Facility, (4) Design development and calibration of instrumentation for the German Upper
Plenum Test Facility (UPTF) entails testing in a "see-through"representation of the UPTF at Oak Ridge. A similar facility is usedat EG&G. (5) A video camera can be coupled to the EG&G Imaging Optical Probe to record the flow of high-temperature andpressure steam-water in nuclear reactor piping.

233

Engineering Center at Canoga Park, California.
Results appeared to confirm the conservatism of thepresent model (Le., that the flow through the valve isgreater than predicted in the NRC analysis model).
However, tests may still be needed.

Program Support
As part of NRC's participation in the 3D international program (see p. 199, 1978 Annual Report),
researchers at the Idaho National Engineering
Laboratory and the Oak Ridge National Laboratory
began the loan of advanced instruments to West German and Japanese counterparts for use in their facilities, and scientists at Los Alamos began using the
TRAC code to analyze those test facilities. During
1979, the Japanese completed the large 2000-rod electrically heated Cylindrical Core Test Facility, andcheckout tests have been completed. Work continues
in Japan on the Slab Core Test Facility, expected to be
operational in 1980. Both facilities will provide improved understanding of steam/water behavior in asimulated nuclear core. In Germany, researchers continued design work on the Upper Plenum Test Facility, which will experimentally model the behavior ofsteam and water droplets in the upper plenum of a
PWR during refill and reflood stages of a postulated
LOCA. This cooperative program provides NRC with
better understanding of the physics of accident
behavior and an advanced accident-prediction code
(TRAC) at about one-third the cost of doing the whole
job alone.

Technical Support
Under the Technical Support Program, NRC sharessponsorship with the Department of Energy of the
Nuclear Safety Information Center at Oak Ridge andthe National Energy Software Center at Argonne inIllinois.
Nuclear Safety Information Center. The Nuclear
Safety Information Center (NSIC) at Oak Ridge provides a focal point for safety information on reactors
and other nuclear facilities. Technical experts who are
cognizant of the literature in each area of specialty
provide replies to questions from NRC, DOE and thenuclear community. Information is provided to nonexempt customers on a cost recovery basis. Seventeen
reports, in addition to the bimonthly review, Nuclear
Safety, were published in 1979. The NSIC also gavesignificant support to the ACRS, and its consultants
during its review of the Licensee Event Reports.
National Energy Software Center. The National
Energy Software Center (NESC) at Argonne National
Laboratory is partially funded by NRC to make NRCsponsored computer codes available to the public. Between October 1978 and September 1979 the Center
distributed 1180 copies of the software packages in

response to requests from NRC and DOE offices, theNuclear Energy Agency Data Bank in France, other
U.S. government agencies, universities, and commercial and industrial organizations. On September 30,
1979, the NESC list of software packages available for
distribution contained 45 items (codes) specificallysponsored by NRC.
Water Reactor Safety Information Meeting. The
NRC held its sixth Water Reactor Safety Research Information Meeting November 6-9, 1978, at the National Bureau of Standards, Gaithersburg, MD. One
hundred sixteen papers were presented, including 12
invited papers dealing with foreign safety research
programs. More than 700 persons attended including
175 foreign visitors.

Advanced Reactor Research
NRC's Advanced Reactor Safety Research program
focuses on two reactor concepts: High Temperature
Gas-cooled Reactors (HTGR), and Liquid Metalcooled Fast Breeder Reactors (LMFBR). Fiscal year
1979 activities in each of these areas are outlined
below.

HIGH TEMPERATURE GAS-COOLEDREACTORS
The President's fiscal year 1980 budget eliminatedfunds for gas-cooled reactor research, and NRC programs in that area were targeted for termination in thefirst several months of that year. Some tasks had been
discontinued by mid-1979. Four DOE laboratories
were affected by the terminations. To assure an orderly termination and provide all possible allowances forpossible future resumption of the research, guidelines
were provided for actions in which: (1) items of particular significance to the operation of the Fort SaintVrain reactor (FSV) in Colorado were identified; (2)distinctions were made between programs which
would be costly to discontinue in terms of data-loss,
restart expertise, etc., and those which could beresumed fairly easily, and (3) the impact on contractor
personnel would be minimized, with assurances that
cadres of expertise can be maintained at eachlaboratory in case of a decision to resume the research.
Within these guidelines, a number of programs could
be carried out, although at very low levels of activity.
Some examples: the metals and graphite programs
were kept going at Brookhaven National Laboratory;transient analysis and seismic core modeling of Fort
Saint Vrain continued at Los Alamos; low-level efforts
continued at Oak Ridge on FSV-related heat transfer,and at Battelle's Pacific Northwest Laboratory (PNL)
on graphite inspection techniques. Other than thesegreatly curtailed activities, however, NRC research inadvanced reactor concepts will be confined to thedevelopment of computer codes and models for future
use in safety investigations.

234

LIQUID METAL-COOLED FASTBREEDER REACTORSThe LMFBR program is subdivided into five areas:
analysis, safety test facilities, materials interactions,aerosol release and transport, and systems integrity.
Progress during 1979 in each of these sub-programs is
discussed below.

Analysis Program
Most of the work in this area was performed at
Argonne, Brookhaven, Los Alamos, and Sandia
Laboratories, as follows:
Argonne National Laboratory completed ananalysis of critical experiments dealing wth LMFBR
safety using the VIM Monte Carlo code and the Zero
Power Reactor-9 (ZPR-9) facility. Results should aid invalidating the neutronics computer codes used in
LMFBR accident analysis. Other ANL code work
resulted in improved calculational efficiencies (by factors up to five) using the COMMIX (Component Mixing) and BODYFIT·1 (Boundary-Fitted Transformation) codes. ANL work in the cooperative studies withEURATOM and the United Kingdom featured
calculations with the SAS3D/EPIC code to quantify
consequences of various accident phenomena, as well
as a study of fuel-pin behavior.
At Brookhaven National Laboratory, work on the"Super System Code" (SSC) continued during 1979(see p. 172, 1977 NRC Annual Report), and a version
modeling the Fast Flux Test Facility reactor (FFTF)
was completed. Startup tests planned for the FFTF
were precalculated and will be compared with
operating data when it becomes available next year.
Los Alamos Scientific Laboratory's analysis program on hypothetical core disruptive accidents(HCDA) in breeder reactors has been shifted fromrapid, energetic accidents to those that develop more
slowly. A key concern in such hypothetical accidents isthe transition phase in which the core begins to melt
and core materials begin to move. LASL completed
the first consistent analysis of the transition phaseusing the SIMMER computer code. SIMMER combines calculations of neutronics, fluid dynamics,
thermodynamics and the interactions of these factors
with one another during an accident. Also the portionof the code that calculates neutronics phenomena wascompletely revised and the revised code was tested
against experiments performed at Argonne. The fluid
dynamics and thermodynamic models of SIMMER
had been tested earlier with generally good results.
The code was made available in 1979 for use in the
United Kingdom, Germany, and the European
Economic Community Research Center in Italy.
Sandia Laboratories completed a preliminary version of a computer code called CONTAIN, for use inanalyzing the responses of advanced reactor containment systems to postulated accident threats. The
code will compute the structural and radiological interactions when core material drops from a primary
reactor vessel onto the containment floor, and willassess the character of the residual mass.
The Univer~ity of Arizona completed the BRENDA
code used for dynamic similation of transients in loop
type LMFBRs.

Safety Test Facilities
Following the upgrading of the Annular Core
Research Reactor (ACRR) at Sandia Laboratories,
NRC's safety test facility work has consisted of installing a new diagnostics system in ACRR, and of implementing the ACRR-CABRI collaboration in fastreactor safety experiments. (Both activities were
described in the 1978 NRC Annual Report, p. 202.)

Materials Interactions
Experiments and analytical model development on
the energetics of severe accidents and on the meltthrough potential of post-accident core debris continued at Sandia Laboratories in 1979. Analysis ofprevious ACRR experiments on fuel pellets under accident conditions showed that rapid fuel swelling
resulted from fission gas production, and this could
not be explained by existing analytical models. As aresult, new models of this phenomenon weredeveloped and an improved ACRR series of experiments on fuel disruption (FD-2) was begun. Experiments on the disruption of irradiated fuel and its
sweep-out from the core by coolant in LMFBR accidents will use the ACRR's new fuel-motion
diagnostics system. A possible spin-off from this work
is a suggested series of tests on light water reactor fuelto examine details of fuel failure in conditions such as
TMI-type accidents.
In Prompt-Burst Energetics work at Sandia, experiments on the damage potential of severe power excursions (prompt bursts) were resumed in the upgraded ACRR. In these experiments an LMFBR fuel pin
contained in a sodium-filled capsule is placed in the
ACRR experiment cavity and exposed to an intense,
short burst of neutrons that melts and may even partially vaporize the fuel. The resulting pressure andmechanical damage potential are measured and usedin constructing analytical models for assessing the'threat of such power excursions to the integrity ofreactor vessels and piping. The experiments are showing considerably less damage potential than previously
considered possible.

Aerosol Release and Transport
Tests of sodium/uranium oxide aerosols in the
Nuclear Safety Pilot Plant (NSPP) at Oak Ridge (see p.

235173, 1977 Annual Report) were directed toward areas
recommended by NRC's Advisory Committee on Reactor Safeguards. An extensive matrix of such aerosols
was examined in a wide variety of conditions in thefirst phase of the test program. It is scheduled to continue through most of fiscal year 1980. The NSPP test
results are used in the assessment of the HAARM-3
aerosol transport computer code. The code assessment
is being performed by Battelle Columbus laboratories
with the assistance of the University ofMissouri-Columbia.

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100

400

1000

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TIME (min)Mixed Oxide Aerosol Concentration - Run 301

As part of the Aerosol Release and Transport program at Oak
Ridge, uranium-oxide (U30.) aerosols and sodium-oxide (Na.O)
aerosols are mixed to permit study of their interactions andbehavior as a function of time. (The U3 0. aerosols are producedby burning uranium in a plasma torch and the Na.O aerosols bybuming sodium in pools or sprays.) Mixing in the Nuclear Safety
Pilot Plant (NSPP) permits study of behavior under conditions inwhich both aerosolS coexist in the secondary containment and arep'resumed to act together. The experiment diagramed here involved
the introduction of NaaO aerosolS into an existing concentration of
UaO. aerosol, causing a marked increase in the U3 0 a removal rate
due to gravitational settling. The ex~riment tended to confirm
that sodium oxide aerosols decrease the concentration of nuclear
fuel (uranium) aerosols in a vessel under postulated LMFBR accident conditions.

Another study at ORNL involves the transport of
U0 2/sodium aerosols through overlying sodium. This
study uses the Fuel Aerosol Simulant Test (FAST)
facility, and trial tests using water instead of sodium toestablish facility characteristics were under way at theend of 1979. Analytical assistance is provided to thisprogram by the University of Virginia.

Systems Integrity
Debris-bed behavior modeled in the first three inreactor tests at Sandia Laboratories, using the AnnularCore Research Reactor (see p. 176, 1977 Annual
Report and p. 204, 1978 Annual Report) was used incoolability analyses of the Three Mile Island accident.
Following that work, a fourth test (of a planned
16-test matrix) dealing with cool ability as a function of
sub-cooling was performed, with results indicatingthat self-rearrangement of the debris bed enhances its
coolability.
A special series of large-scale sodium concrete interaction tests was completed in support of the NRC's
final safety evaluation of the Fast Flux Test Facility(FFTF). The results of the test confirmed the staff
position regarding FFTF containment margins.
Another test program was initiated to study theinteraction of molten fuel materials with candidate
materials that could be used in place of concrete tocontain core debris from a postulated core meltdown.
A large fuel melt test facility is under construction at
Sandia. The facility will be used to conduct tests containing up to one half ton of molten fuel in contact
with structural materials to confirm analyticalmethods for predicting containment system response
under postulated accident conditions.

General Reactor
Safety Research
NRC's General Reactor Safety Research comprises
three areas: site safety research, mechanical engineering research, and structural engineering research.
These were described in the 1978 NRC Annual Report(see pp. 206-208).

SITE SAFETY RESEARCH
Site safety research is generic research directed
toward estimating the effects on nuclear facilities ofearthquakes, floods, and tornadoes and other severe
phenomena, understanding the distribution of those
severe natural phenomena, and providing information
on meteorology affecting the atmospheric dispersion ofradionuclides under postulated accident conditions.

Geology and Seismology
A magnitude-4 earthquake occurred about six miles
from the Maine Yankee nuclear power plant near

236

Wiscasset, Maine on April 18, 1979. Following thatevent, Boston College and the Maine State Geologist's
Office cooperated in installing a dense array of portable seismographs to record aftershocks which would
locate the causative fault. The aftershock pattern suggested a north-south trend clustered around theepicenter of the main shock. The last aftershock
was recorded on June 21, 1979.
In the Charleston, S.C., region, high-precision vertical seismic reflection profiling has revealed a probable fault in deep subsurface rock near the center ofthe large 1886 earthquake. (See p. 206, 1978 Annual
Report.) This is the first direct evidence of a possible
causative geologic structure for that earthquake.
Studies are continuing to determine the extent, historyof movements, and tectonic relationships of the fault
in order to assess its potential earthquake hazard.
In other 1979 activities under this program:• A new arrangement for direct cooperation with
the Canadian Department of Energy, Mines andResources will add 12 high quality seismographstations to "look" southward into areas of interestin the Northeastern U.S. This also will improve
U.S. capabilities to evaluate earthquake regions
along the St. Lawrence River and in the northwest extension of the problematic "BostonOttawa seismic trend.
• Summaries and interpretations of known data
bearing on earthquake hazard assessment in areasof the Northeastern U.S., in the New Madrid,Mo., region and in the midcontinent region were
published during 1979.
• Studies of the response of soil foundations toearthquake motions are important to earthquake
design of power plants and other structures. In
1979 NRC-supported studies of the foundations ofimportant accelerograph stations were completed. Other geotechnical studies resulted inpublication of a technical manual describing
equipment and operations for determining
dynamic soil properties in place.

Meteorology and Hydrology
Projects in this research field included the following:
Severe Storms. Damage surveys of the December 4,1978 tornado that struck Bossier City, La. and the tornado that devastated Wichita Falls, Tx., on April 10,
1979, were performed. Six 750-pound wide-flange
steel beams, 18- and 24-ft. long, were hurled up to 300
yards in the -Bossier City tornado, two of thempenetrating the ground about eight feet. The mostsignificant aspect of the Wichita Falls tornado was its
size-up to one mile wide and more than 40 miles
long. Information compiled from these surveys provide authoritative data against which to evaluate thedesign criteria developed for nuclear power plants andfuel cycle facilities.
Flooding. A research program was initiated in 1979
to quantify the safety margins used in flood~related
design criteria for nuclear power plants. Flood probabilities, as a function of geographic location, andwith particular reference to coastal phenomena, will
be determined. A numerical simulation of theNovember 1975 tsunami in Hawaii was completed
during the year, and the storm surge and wave heightassociated with the passage of Hurricane David along
the east coast of Florida in September were measured.
Atmospheric Diffusion. The NRC-supported atmospheric dispersion research program featured fullscale field tests and wind tunnel simulations ofbuilding-wake-dispersion characteristics; the use of
gaseous tracer and lidar technologies to measure vertical diffusion over different terrains, and determinations of thermal performance of cooling and fixedspray ponds used as heat sinks. Planning also beganlate in the year for a comprehensive field and modeling program to study atmospheric diffusion in a complex shoreline environment.

MECHANICAL ENGINEERINGRESEARCH
NRC's new Mechanical Engineering Research Program, initiated in 1978, provides the licensing staff
with improved methods for evaluating the safety andstructural integrity of systems, components, andequipment under normal and accident conditions interms of margins of safety and probabilities of failure.
Major sub-programs are:Seismic Safety Margins Research Program (SSMRP).
This multi-discipline program at Lawrence Livermore
Laboratory is designed to estimate the conservatisms
in the seismic safety requirements stated in the NRC
licensing standard review plan, and to improve thoserequirements. The approach is to develop a probabilistic methodology that can realistically estimate
the behavior of buildings and components of a nuclear
plant during an earthquake. The first phase of the program will be completed in 1980.
Nonlinear System Modeling Program. A simplified
computer code for the analysis of piping systems was~ompleted; a mathematical model of a simplified
mechanical system (typical of those in a nuclear plant)
was validated; and design charts were issued for use indescribing the motion of mechanical equipment, piping and components. There is a need for further
research to better characterize the dynamic responseof valves in nuclear plants and to better model andscale mechanical systems and equipment.
PARET Program. PARET is a systems identification
technique to determine frequencies, mode shapes and==================================================237
damping values for complex systems. The design ofbasic techniques and a computer program were completed during the year. Some test plans and recommendations for experiments to aid in determining certain dynamic parameters of nuclear power plants also
were developed. These parameters will be used to confirm safety evaluations for earthquakes, blowdowns,
and other accident or environmental events.
Piping Benchmarks. Several piping benchmarks
were developed at Brookhaven National Laboratory.
Benchmarks are used to validate the computer programs used in the dynamic analysis of power plant
piping systems. Some of the benchmark problems fromof this project were used in evaluating five nuclear
power plant shutdowns (see Chapter 3) which occurred in March 1979 when errors were discovered in theanalytical computer codes. Licensee codes were checked against benchmark problems, and agreement between them was required before the plants could
resume normal operation.
Load Combinations. A probabilistic study dealing
with load combinations was undertaken in 1979 which
will provide guidance on design requirements for
simultaneous occurrences of LOCA's and earthquakes,
and will develop dynamic response combination
methodologies.

STRUCTURAL ENGINEERINGRESEARCH
Structural Engineering Research is generic research
to assess the safety of nuclear plant structures subjected to extremes of natural events in all possible
operating and accident conditions. Primary areas of
exploration during 1979 were the definition of design
loads on nuclear plant structures and the behavior ofsuch structures. Individual projects in these areas are
the following:
Soil-Structure Interaction (SSI). A simple computercode for the evaluation of SSI has been completed.
Two other codes will permit independent evaluationby the licensing staff of SSI analyses submitted by applicants. These appear to promise a reduction in licensing time heretofore spent verifying licensee calculations.
Seismic Design Criteria. This program deals with anunresolved safety issue for which research support isprovided to the licensing staff (see Chapter 3). Reportsof research projects on the quantification of seismic
conservatisms, elasto-plastic seismic analysis, site
specific spectra, seismic input and SSI, nonlinear
structural dynamic analysis procedure have been completed; Studies in the modeling of earthquakes andDevelopment of mathematical models to represent the response ofreactor com~onents to potential accident conditions begins with actual tests ana simulations. Here a nuclear-type valve is subjected tovarious patterns of seismic shock excitation to permit researchers toderive dynamic response and behavior characteristics for code use.
The "shake table" is located at Hughes Aircraft Company, Fullerton, Califomia.

analyses of near-field ground motion continued at theend of the year, and recommendations on modifying
seismic design criteria were almost finished.
Seismic Shear Transfer. This comprehensive program encompasses: scale model testing at Cornell
University; full scale tests of wall segments at thePortland Cement Association; and analytical efforts at
the Massachusetts Institute of Technology, to provide
a better basis for the safety assessment of containmentbuildings. Several tests and analyses had been completed by year's end.
Methods of Seismic Qualification. A report,
NUREG/CR-0345, "An Evaluation of Seismic
Qualification Tests for Nuclear Power Plant Equipment," compared the results of a series of six tests ofthe performance of a typical nuclear power plant electrical cabinet in seismic events. Improvement of thecriteria defining vibratory input was recommended.
Water Hammer Effects. This research evaluates theeffects and safety significance of water hammer
phenomena in nuclear power plants. A technicalreport, NUREG-0582, "Water Hammer in Nuclear
Power Plants," published in July, provides the results

238of a staff review of water hammer events and statescurrent staff licensing positions on the topic.
In addition, seven technical reports on water hammer were issued. Together they provide a review andevaluation of actual and potential water hammer
events in nuclear power plants, analytical methods
and calculation procedures to be used in evaluatingwater hammer incidents, and state-of-the-art information on water hammer. (See Chapter 3, "Water Hammer.")
Tornado Generated Missile. Research· in this area
was oriented in 1979 to automobiles, consideredamong the objects most likely to become airborne andthreaten nuclear power plants in a tornado. Current
practice in impact calculation assumes that the
automobile is a rigid mass and this results in the imposition of ultra-conservative design requirements.
NRC's reoriented tornado missile project examines the
behavior of automobiles at high-impact velocities, anapproach which promises more realistic plant design
reqUirements.

Fuel Cycle, Environmental
and Waste Management
Research
NRC's Fuel Cycle, Environmental and Waste
Management research aims to confirm the basic dataand predictive models used in assessing safety in theroutine operations of reactors, fuel cycle facilities,
transportation of radioactive materials, and disposalof radioactive wastes.

FUEL CYCLE RESEARCH
Fuel cycle research in 1979 focused largely on commercial operations associated with the milling of
uranium ore, the fabrication and storage of powerreactor fuels, and the transportation of radioactive
materials. Earlier research in the uranium fuel recycle
area was canceled as a matter of national policy. The
following were 1979 highlights:• Experiments to determine the rate at whichuranium oxide, or "yellowcake" dissolves insimulated human digestive juices and lung fluids
indicated that it dissolves more rapidly thanpreviously believed. A follow-on study was
undertaken to verify these results in rats, dogs,
and monkeys.
• The effectiveness of so called "high efficiency"
filters used to reduce airborne releases to the environment was tested under simulated tornado
conditions. While the filters did not survive thetest conditions of NRC Region I tornadoes (maximum differential pressure = 3 psi), they do survive reduced pressures and air-flows. This information will be useful in designing future plantventilation systems.
• Criticality experiments provided benchmark data
related to fuel-element storage and shipping configurations. Data from these experiments were used to validate NRC methods of analyzing licensee
criticality safety programs.
• Transportation research programs were directed
toward providing verified codes for analyzing
damage to large shipping packages during transportation accidents. At the end of the year puncture resistance assessments had been completedusing laminated plates to represent large shieldedshipping cask end-plates, and a computer model
to assess the shocks experienced by shielded casks
during normal rail transport was being checked.
• A new program was developed to establish the
response of spent-fuel casks to acts of sabotage involving explosive threats and the potential consequences of such acts on the public.

ENVIRONMENTAL RESEARCH
Environmental Effects Research is concerned withthe radioactive chemicals and heat which are released
from nuclear-power plants, with their movement
through the environment, and with improving
methods of predicting their impacts on people and theenvironment-impacts which include not only healthand safety, but the social and economic effects from
nuclear power plants, as well.

Radiation Dosimetry and Health Effects
A method for assuring that occupational radiation
exposures at nuclear power reactors are kept as low as
is reasonably achievable (ALARA) was developed anddocumented in 1979 with the 'issuance of
NUREGI CR.0446, "Determining the Effectiveness of
ALARA Design and Operational Features." In addition, a study of animals continuously exposed to lowlevels of gamma radiation has led to new research todetermine if an animal's susceptibility to radiationinduced leukemia is affected by genetic disorders. This
program may ultimately result in better assessment oftHe effects of human exposure to radiation.

Ecological Impact Studies
NRC is investigating the effects of chlorine compounds released from power plants on important fish
species. Chronic toxicity studies to date have indicatedthat rainbow trout are not adversely affected by theamounts of chlorine normally discharged from nuclear
stations, that chlorine byproducts (chloroform,

239

bromoform, etc.) have little impact on shrimp,
oysters, and other fish species. A mathematical model
developed in 1979 to predict concentrations ofchlorine compounds in discharged cooling water permits a more precise estimate of their effects. Validation of the model, using field data from an operating
nuclear station, will be undertaken in 1980. Other
studies at nuclear power stations have shown that copper is released to the environment during plant operations, and measurements of the effects of this metal on
various fish were undertaken at coastal and estuarine
sites toward the end of the period. NRC has also investigated the release of asbestos fibers to the environment from power plant cooling towers, and found thatthe water from cooling towers which have asbestos-fillcontains asbestos fibers. This was determined to be of
no immediate significance to the public health but
worthy of continued monitoring.

Environmental Transportand Effluent Monitoring
Field studies were completed at the West Valley,
N. Y., waste burial site to validate a transport model
simulating radionuclide movement in river systems.
The studies included measurements of data on channel
characteristics, water flow rates, and radionuclide
concentrations in Cattaraugus and Buttermilk Creeks
extending from West Valley to Lake Erie.
Other studies were done to measure the levels ofradioiodine, carbon-14, and tritium in the environment adjacent to the Quad Cities nuclear station.
Measurements thus far show no detectable quantitiesof tritium or carbon-14 within 5 kilometers from thesite. Trace amounts of radioiodine, xenon, and krypton were measured at the same locations, andradioiodine transport mechanisms are being evaluatedfrom those measurements. Follow-on studies are in
progress on radioiodine behavior.
Independent NRC measurements of radioactive
materials in liquid and gaseous effluents at four
operating PWRs (Fort Calhoun, Zion, Turkey Point,
and Rancho Seco) show generally good agreement
both with earlier measurements made by plant
operators and with estimates by the NRC staff. This
information is used by NRC to assure that releases aremaintained at levels as low as is reasonable achievable.
The effects of atmospheric pollutants such as nitrogenand sulfur oxides also have been measured, and thedata obtained were being evaluated for use inestablishing filter replacement schedules for nuclear
plants.
To support the development of NRC standards forreactor decommissioning, research has been initiatedto determine both the sources of long-lived radioactive
products in structural and shielding materials and thenature and distribution of radioactive contaminants
within such plants. The costs and methods of removing
such contamination also will be studied.

Socioeconomic Impacts
and Regional Siting
A method for obtaining quantitative estimates of thevisual impact of different nuclear power plant coolingsystems have on communities has been completed, andthe results may enhance NRC's ability to weigh thecosts and benefits of alternative cooling systems.
NRC also undertook to develop a modeling system
to assist State authorities in assessing of environmental
impacts and siting alternatives. The project, now in
Phase II in New England, consists of a regional electricenergy demand model, a power-generation mix
model, and a power facility siting model. When completed, the system will give State regulatory bodies access to the quantitative estimates of energy/ environment tradeoffs and other information needed for informed siting decisions.
A project begun in 1978 to study the socieconomic
impacts of the construction and operation of nuclear
power plants was modified and expanded during 1979
to assess the impacts of the Three Mile Island accident.
Data input to the study from Three Mile Island will include economic costs (evacuation costs, loss of production, and costs to local governments), the sociologicalimpact of the incident, and the effect of the incident
on land use and land values.

WASTE MANAGEMENT RESEARCH
NRC's waste management research program was set
up to establish an independent data base for licensing
decisions on high-level waste repositories, shallow
land burial sites, and uranium-mill tailing operations.

High-Level Radioactive Waste ResearchHigh~evel waste research in 1979 was divided into
the Material Science Programs and the Geotechnicaland Sciences Program:Material Science Programs. Quantitative relationships between solid (glass-like) waste forms and alimited number of underground environmental andother parameters were investigated in 1979. The
effects of temperature, pressure and chemical influences on interactions between waste forms and surrounding media were also investigated and a
preliminary analytical model was developed forpredicting the long-term performance of silicate
glasses.
Geotechnical and Sciences Program. This program
deals with the molecular movement of dissolved solids;
flow of ground water; fundamental concepts of flow in
fractures (as opposed to flow between grains);
development of numerical predictive models; and theresidence time of ground water. The potential escapeof waste materials is being addressed through research

240================================================
into the sealing of construction-induced openings andnatural fractures using man-made materials. Indirect
determinations of the geologic structure of waste
burial site media by geophysical methods are being
studied toward minimizing the need for exploratory
physical excavation or core boring in such rock.

Low-Level Radioactive Waste ResearchStudies continued in 1979 to assess the migration ofwastes at the shallow land burial sites at Maxey Flats,
Kentucky; West Valley, New York; and Sheffield, Illinois. At the end of the period, radioactive andchemical substances in trench water had beencharacterized and potential migration pathways werebeing analyzed. A laboratory test program on soil
retention of radionuclides and the uptake of ra~
dionuclides by agricultural crops is being extended toinclude field studies at the Maxey Flats site. These arebeing done in coordination with a complementary
program of the U.S. Geological Survey. Results will be
used to assess the applicability of the laboratory studies
to site characterization and predictions of waste
mobility. Such information is needed in making decisions about decommissioning existing burial sites andto improve criteria for the design, operation andmonitoring of future low level radioactive waste
disposal facilities. Tests of the characteristics of solidified low-level wastes were described in the NRC
topical report, "Properties of Radioactive Waste and'Containers" (NUREG/CR-OBI9).

Uranium Milling Research Program
A phase of the field research program at operatingmills to support licensing activities and the development of the "Generic Environmental Impact Statement on Uranium Milling," was completed and thereport (NUREG-0511) was published in April 1979.
The program included field tests to characterize thenature and extent of airborne contaminants at active
mills and to assess their potential impacts on air andwater resources. In 1980, the program emphasis will
shift to the development of information relative tosuch mitigative measures as groundwater protection
and radon attenuation.

Risk Assessment Research
NRC risk assessment research embraces the development of methods and data for probabilistic nuclear
safety analysis, reliability analysis, and the predictionof risks. Activities during 1979 included the development of improved techniques to predict nuclear accident consequences; reactor risk assessment and licensing support; fuel ~ycle risk assessment, and thedevelopment of statistical methods and data-bases
necessary for risk assessment.

Lewis Group Review of W ASH-1400
At the time of publication of the 1978 NRC Annual
Report, the Commission was studying the report of a
special review group headed by Professor Harold W.
Lewis of the University of California which waschartered to: (1) clarify the achievements and limitations of the Reactor Safety Study (WASH·1400, issued
in 1975, also known as the "Rasmussen Study"); assess
peer comments on it, and the response to those comments; (2) study the present state of risk assessment
methodology; and (3) recommend how and whether
such methodology can be used in the regulatory process. In general, the Lewis report agreed with much of
the criticism that had been expressed of the Reactor
Safety Study, particularly of the Executive Summaryof the study, while endorsing the basic fault tree/ event
tree methodology that was employed in the study. (See1978 NRC Annual Report, p. 213, for summary ofLewis report finding.)
In January 1979, the Commission issued a policy
statement accepting the findings of the review group.
The Commission, among other things, withdrew any
explicit or implicit past endorsement of the Executive
Summary of WASH-1400, which the Lewis group concluded had lent itself to misuse in the discussion ofreactor risks; agreed that the peer review process
followed in publishing WASH-1400 was inadequate;
and accepted the Lewis group report's conclusion thatabsolute values of the risks presented in WASH·1400
should not be used uncritically either in the regulatory
process or for public policy purposes. In particular, inlight of the review group's conclusions on accident
probabilities, the Commission said it does not regard
as reliable the Reactor Safety Study's numericalestimate of the overall risk of reactor accidents. The
Commission noted, however, that it supports the extended use of probabilistic risk assessment inregulatory decision-making, and provided the staff
with additional detailed instructions on use of risk
assessment techniques and results.
The events at TMI showed serious problems in many
areas that were noted as problems in WASH.1400,
namely, transient events, small LOCA response, andhuman error. This emphasized the need to use thevaluable tool of probabilistic risk assessment in theregulatory process. Inquiries into the TMI accident
have also urged NRC to do this.
Probabilistic risk assessment techniques are now being used much more widely in the NRC. They wereused recently, for example, to compare generic safetyissues according to their contribution to risk, and thishas allowed the Commission to focus its efforts on
those 18 issues which pose the highest risk. The remaining 115 items will be addressed later (see"Unresolved Safety Issues," in Chapter 3).

================================================241
Reactor Accident
Conseq uence Analysis
A program to develop a site-specific consequencemodel recommended by the Lewis group was undertaken to update the Calculations of the Reactor Accident Consequences (CRAC) model, developed for theReactor Safety Study. Improvements, including more
realistic treatment of population movements andemergency response and a new meteorological sampling technique, should give a more realistic predictionof the consequences of accidental releases of radioactive materials at specific locations. When completed,the consequence model will be able to differentiate
between acceptable and unacceptable sites based on
public risk criteria.

Reactor Risk Assessment
And Licensing Support
NRC continued in 1979 to expand the application ofrisk methodology to a broader spectrum of LWR safetyissues, and to apply the methodology and related
engineering insights to issues of immediate concern to
the licensing staff. This included special efforts toanalyze the accident at Three Mile Island.
Activities included: sensitivity studies on potential
core meltdowns, (including the Three Mile Island accident) to aid in setting priorities for meltdownaccident research; applying fault-tree and event-tree
methodology to other LWR design concepts tobroaden engineering insights; evaluating risks in L WR
accidents that do not lead to core melting; assessing
risks to the public from radioactive contamination of
the hydrosphere as a result of core-melt accidents; andassessing the impact of external events (such astransportation aCcidents) on nuclear plants. In the
category of reactor licensing support, a ranking ofgeneric safety issues from a risk perspective waspublished (see "Identification of Unresolved Safety
Issues Relating to Nuclear Power Plants,"
NUREG-0510, and p. 19, 1978 Annual Report).
Analyses related to Three Mile Island included acooperative effort with licensing engineers to evaluate
and improve the reliability of auxiliary feedwater
systems in operating nuclear plants. Other efforts included an event-tree analysis and recommendations todeal with anticipated-transients-without-scram (seeChapter 3); a value/impact study of the standard
review plan for BWR's, and continued evaluation andrecommendations for the use in licensing of test intervals for certain plant components such as valves andpumps. Probabilistic assessments of direct current
power supplies and of protective measures against lossof alternating current power were also initiated.

Fuel Cycle Risk Assessment
The objective of fuel cycle risk assessment is to identify the important contributors to risk from nuclear
fuel cycle activities, other than reactors, using conseqll:ence and probability models. In 1979 a major effort
usmg such models was made at Sandia Laboratories todevelop risk methodology to examine deep geologic
isolation of high-level radioactive waste in bedded
salt. Work in 1980 will focus on licensing questions
and on expanding the methodology on other isolation
media. Several documents on waste isolation risk
methodology were published: "Risk Methodology for
Geologic Disposal of Radioactive Wastes: Interim
Report" (NUREG/CR-0458), "Risk Methodology for
Geologic Disposal of Radioactive Waste: Sensitivity
Analysis Techniques" (NUREG/CR-0394), and "Risk
Methodology for Geologic Disposal of Radioactive
Waste: The Sandia Waste Isolation Flow andTransport (SWIFT) Model" (NUREG/CR-0424). In
order for the NRC staff to build up expertise in the useof the risk methodology, the Interoffice Waste
Management Modeling Group (IWMG) was formed.
By documenting the insights gained from exercising
the models on a series of increasingly complex waste
repository assessment problems, the IWMG work is
designed to improve understanding of the strengths
and weaknesses of the risk assessment methodology,
and to assist in the licensing decision process.
Planning was completed for a three-year fuel cycle
project ,review to obtain a balanced, independent,
multidisciplinary appraisal of results from the wasteisolation risk methodology program. Other 1979 activities included initial steps to develop methodologies
for assessing spent-fuel isolation risks and for examining the management of certain radioactive gases emitted from fuel-cycle facilities. The Environmental Protection Agency is investigating the use of probabilistic
risk analysis in its regulatory activities, and the NRC
staff and consultants reviewed EPA-sponsored work in
the assessment of waste isolation. The review group
recommendations are being implemented in both NRC
and EPA risk methodology programs.

Methodology Development
NRC programs to develop methodology for probabilistic safety analysis and risk assessment deal withthe hazards to nuclear plant operations from fires andfloods, and with the impact of testing schedules on thereliability of engineered safety features. New initiatives in fire and flood risk assessment were taken in
1979. The impact of testing schedules on the reliabilityof safety systems is studied using a computer program
named FRANTIC (see p. 180, 1977 NRC Annual
Report). In 1979 the program was extended and improved to better portray time-dependent effects on
system reliability.

242

Programs for the collection and analysis of statistical
data initiated in 1979 addressed the reliability of
nuclear plant operators, maintenance personnel, and awide variety of safety-related components. Data on
equipment failures and human errors were drawn
from licensee event reports, power plant log books,
and the records in the Nuclear Plant Reliability Data
System data banks. The data were analyzed to estimate average error or failure rates; check the consistency of the data sources; identify trends or patterns
in the data (time trends, variations from plant toplant, etc.), and study multiple failures of common
cause. Work continued on a handbook describing theprincipal factors governing the reliability of nuclear
plant operators to serve as a guide to safety analysts
and risk assessment practitioners in assessing humanerror contributions to risks.

Research to Improve
Reactor Safety
An amendment (P.L. 95-209) to the Energy
Reorganization Act of 1974 directs NRC to "develop a
long-term plan for projects for the development of new
or improved safety systems for nuclear power plants"
and requires that the plan be updated annually andsubmitted to the Congress. The Congressional intent
behind this effort is "the improvement of reactor safety
and not the enhancement of the economic attractiveness of nuclear power versus alternative energy
sources.
In April 1978 NRC submitted to Congress a "Planfor Research to Improve the Safety of Light-Water
Nuclear Power Plants" (NUREG-0438), which
presented an evaluation of concepts proposed to improve safety and recommended a three-year, $14.9
million research program. The objectives are to determine the feasibility of achieving particular improvements in safety, to evaluate the safetysignificance of proposed changes and to propose
regulatory requirements where implementation isdetermined to be desirable, without preparing detail.
ed designs. Five research topics and two general
studies were suggested:
Alternate containment concepts-especially vented
containments-to mitigate the consequences of
postulated core meltdown accidents. This is accomplished by improving control of the release ofradioactivity to the environment.
Alternate decay heat removal concepts-especiallyadd on, bunkered systems-to reduce the probabilityof core meltdowns by increasing the reliability ofsystems designed to remove heat from the reactor core
after fission ceases.

,Alternate emergency core cooling concepts-todevelop simpler and more clearly demonstrable
systems to prevent fuel overheating in the event of pipe
rupture.
Improved in-plant accident response-to reduce therisk from human error by enhancing the quality of theoperator-machine interface and by helping operators
make correct decisions during accidents.
Advanced seismic design-to reduce the vulnerability of plants to earthquakes by decoupling or
strengthening components against seismic forces.
Scoping studies oj other concepts-to determine
their potential for improving safety and to assess theneed for further research. The studies address protection against sabotage, better ways to monitor the condition of the plant, new siting concepts and ways toreduce occupational exposure without increasing
public risk.
Improved evaluation methodology-to assist in
making more rigorous and thorough assessments of thevalues and impacts associated with these concepts, andin planning future safety research programs.
Operating experience accumulated since
NUREG-0438 was issued, including the events atThree Mile Island in March 1979, reinforces the
judgments expressed therein. Many individuals andorganizations have submitted additional recommendations for improving safety since TMI, and these also
tend to support the original judgments, especially thehigh priorities assigned to improved in-plant accident
response, alternate containment concepts and alternate decay heat removal systems. Within the highpriority areas, research toward enhancing thecapabilities of reactor operators and improving thequality of the operator-machine interface has been accelerated.
In fiscal year 1979 Congress authorized the expenditure of $1,500,000 to implement the research plan,
but appropriated no funds for the purpose. To accommodate this mismatch, the Commission sought andreceived from Congress reprogramming approval in
two separate actions totaling $800,000. The time required to complete these actions delayed the initiationof technical work until the latter half of the fiscal year.
Furthermore, the reduced amount of funds made itnecessary to restrict studies to a few specific concepts
judged to have the highest potential for risk reduction.
These factors are reflected in the limited results
reported here.
In 1979 research focused on improved in-plant accident response, vented containment, and add-on decay
heat removal systems. The status and direction of programs underway at the end of the fiscal year are summarized below.

==================================================243
Improved In-Plant Accident Response
A study was initiated at Idaho National Engineering
Laboratory to identify the information required by anoperator to determine unambiguously the status of hisplant. Accident sequences having relatively high pro~babilities of leading to core damage are being analyzed
to identify how the plant might respond and what
measurements are needed to accurately and uniquely
characterize that response. The study indicates what
range of physical parameters should be measurable
and for which parameters direct measurement might
be preferable to indirect measurement. The results are
helping establish regulatory positions regarding instrumentation required to monitor the course of an accident. They are also useful in determining what data
might be transmitted to remote monitoring and technical support centers.
The simultaneous influx of alarms and data during areactor accident can overwhelm even a well-trained,
experienced operator. A potential solution lies in theapplication of computer technology to the diagnosis ofplant disturbances. Oak Ridge National Laboratory isreviewing the state-of-the-art in computerized
disturbance-analysis systems and audio-visual displayswith a view toward transferring the applicabletechnology to nuclear plants.
Oak Ridge is also examining key aspects of thetechnical basis for disturbance analysis systems, including the reliability of the hardware and softwareand the quality of the plant systems analysis. These insights will enable the NRC to establish regulatory requirements and to evaluate related efforts by the
nuclear industry.
Planning for a conference jointly sponsored by the
NRC and the Institute of Electrical and Electronics
Engineers was initiated in fiscal year 1979 to identify
the extent to which technology from aerospace,defense and other industries may be useful in these
NRC programs.

Alternate Containment Concepts
Risks to the public from nuclear reactor accidents
are dominated by sequences in which the core melts
and the containment ruptures above ground level
because of overpressurization. Concepts have been
proposed which would reduce the pressure in the containment during accidents while channeling the containment atmosphere through filter media to retain
the radioactive materials it might contain.
In a Sandia Laboratories investigation of thetechnical feasibility and risk reduction potential ofthese systems, the state-of-the art in containment
design and filter technology has been surveyed and incorporated into a program plan. Conceptual designs
fitting vent filter systems to several existing containment types are being developed, and computercodes to analyze their effectiveness are being modified

as necessary. The goal is to propose regulatory requirements for vent filter systems and to assess the values
and impacts of their implementation on existing andfuture reactors.

Alternate Decay Heat Removal Systems
After the nuclear chain reaction in a power reactor
ceases, residual heat produced by the decay ofradioisotopes must be removed over an extended
period of time. This is the function of decay heatremoval systems. Improvements in the reliability ofthese systems offer relatively high potential for risk
reduction.
A study was initiated at Sandia Laboratories to
identify ways to enhance that reliability and to quantify the risk reduction potential inherent in such improvements. The study includes review of existing
criteria for the design of decay heat removal systems
and estimates of the resulting reliability. Preliminary
performance and safety design requirements are being
generated to guide the development of detailed designs
by the Department of Energy. The reliability of the
improved system will be estimated and the information used to identify the need for and nature of revised
regulatory requirements.

Improved Methodology
Assignments of priorities for safety research anddecisions on the adoption and implementation ofregulatory pOSitions require judgments regarding thevalues and impacts of alternatives. A study was initiated at Pacific Northwest Laboratory in conjunction
with Battelle Columbus Laboratory to develop moreobjective and precise methods for making value/impact assessments. Initial emphasis is on better ways toquantify the risk reduction potential of proposedchanges in design or operation. Longer term activities
include improving the methods used to address theuncertainty in risk estimates and to integrate risk
reduction potential with other values and impacts
affecting a decision. The results would also pertain to
activities beyond the scope of the improved safety
research, such as confirmatory research and adoptionof regulatory standards.
No work was initiated to study containment concepts other than vented containment or decay heatremoval concepts other than a dedicated add-on
system. No work was initiated to investigate advanced
seismic designs or alternate emergency core coolingconcepts. These are being deferred pending the outcome of confirmatory research on the effectiveness ofcurrent designs. Needs for research on concepts otherthan those identified above (e.g., protection against
sabotage, improved techniques for nondestructive examination, improved emergency planning, reduced
occupation exposure) are being reviewed continually244

by the NRC staff within the context of confirmatory
research and the development of requirements by theregulatory staff. Such needs will be addressed in 1980
and beyond, as resources allow.
In addition to conducting its own research, NRC has
provided guidance to the DOE regarding its efforts toimprove reactor safety, including recommendations
that DOE initiate more detailed examinations anddesign studies on computerized disturbance analysis
techniques, vented containment, hydrogen control
technology, add-on decay heat removal systems andseismic decoupling. In addition, NRC asked DOE totake the lead in assessing the cost impacts of design improvements, ihcluding implications for retrofit. This
cooperation should ease the transition of new concepts
into commercial application.

12
Informing andInvolving the Public
NRC's Harold Denton(left) kept press andpublic informed during the early weeksof the TMI recovery.

Many actions are taken annually by the NRC to inform the public directly and to make information
available regarding nuclear regulation. These take theform of public announcements and Federal Register
notices; publication of staff and contractor reports;
providing access to documents in localities across thecountry; holding meetings and workshops with public,
State and local representatives on issues of widespreadinterest; responding to public and Congressional inquiries; opening to public observation Commission,
staff and advisory committee meetings; and many
public hearings on rulemaking and licensing.
Members of the Commission and the NRC staff also
participate in press conferences and public meetings
and, where special interest warrants, testify before
committees of the Congress.
As the most direct means of communicating to the
public, the NRC issues announcements on a wide
range of topics from headquarters and the five
regional offices to some 5,000 members of the news
media, industry, the scientific community and thegeneral public.

Making Documents Available
The NRC maintains its principal Public Document
Room (PDR) at 1717 H Street, N.W., Washington,
D.C., and has established more than 130 public document rooms throughout the country. The local PDRs
are typically located in libraries in cities and towns
near proposed and actual nuclear plant sites, and contain detailed information specific to the nearby
facilities which are either licensed or under regulatory
review. (See Appendix 3 for a list of all local PDRs.)
NRC currently makes publicly available through its
main PDR in Washington approximately 320 new
documents each day. The main PDR contains about
880,000 documents in hard copy or in microfiche.
These documents pertain to the licensing of source

material, production and utilization facilities, special
nuclear material, transportation of radioactive
materials, nuclear exports and imports, research andtechnical assistance reports, reports on generictechnical issues, rules and regulations, Commission
correspondence, transcripts of Commission meetings,
minutes and reports of NRC's advisory committees andother material relating to the responsibilities andoperation of the Commission.
The PDR services a diverse clientele. About 42 percent of its users are from industries directly affected by
the Commission's activities (utilities, vendors, insurers, manufacturers), 12 percent from companies
peripherally affected, 9 percent from law firms, and37 percent from educational institutions, media, congressional and Federal agencies, public interest
groups, and private citizens.
Members of the public may visit the PDR and examine any document in the facility, which furnishes
reference assistance, copying services, and micro-fichereader/printers. In fiscal year 1979, an on-line
bibliographic retrieval system was designed and installed in the main PDR to im;:rease accessibility to
NRC information. During an average month in thereport period, visitors to the Washington PDR requested access to about 7,000 files. The PDR staff also
responded to an average of 90 letters a month. Morethan 1.7 million pages of documents and 13,500
microfiche cards were reproduced for the public during the year.
NRC publishes a Daily Accession List providing abibliographic description of the documents placed in
the PDR. A copy of this list may be obtained bywriting the Division of Technical Information andDocument Control. The PDR also provides limited
free distribution of press releases and Commission/Board orders and issuances.

246

The NRC's main Public Document Room (PDR) in Washington,
shown here, contains about 880,000 documents and makes
available to the public more than 300 new documents -per day-.
Members of the public are encouraged to visit the PDR, which
offers reference, copying and reader/printer services.

TMI.. 2 Investigation Center. In the aftermath of theaccident at Three Mile Island Unit 2 (TMI-2), Senator
Gary Hart, Chairman of the Subcommittee on
Nuclear Regulation of the Committee on Environment
and Public Works, initiated an investigation into theevent. To support this special investigation, Senator
Hart requested NRC to provide facilities for access to
all the documentation related to the accident andother information concerning other operating nuclear
power plants of Babcock and Wilcox (B&W) design.
On June 2, 1979, the Three Mile Island Documentation Investigation Center was established at the Commission·s Washington, D.C., office. The NRC·s Division of Technical Information and Document Control
collected the documentation and staffed the center,
which was also utilized by staff of the President's Commission on the Accident at Three Mile Island.
A pre-incident and post-incident file was established
for all documentation concerning Three Mile Island
Unit 2, including letters, memoranda, technical andsafety analysis reports, pictures, engineering drawings, and depositions taken by investigators. Tapes of
conversations between the NRC Operations Centerand the TMI site and interviews of key participants
were also made available.
The related information on the operating B&W
facilities (Arkansas Nuclear One-Units 1 and 2, Crystal
River, Davis-Besse, Rancho Seco, Oconee Units 1, 2,

and 3, and Three Mile Island Unit 1) included
Preliminary Safety Analysis Reports, Final Safety
Analysis Reports, Environmental Reports, Safety
Evaluation Reports, inspection reports, reportable occurrences, monthly operating and annual environmental reports. A complete set of B&W topicalreports was also available in microfiche. Reference information included NRC Regulatory Guides, thenuclear power Standard Plan, and the results of extensive computer-generated bibliographic searches doneon the Lockheed, RECON, and other energy databases.
A sophisticated array of document management
technology was made available to provide the investigators with prompt access to the large volume of
documentation, which exceeded 2,000,000 pages. The
Commission provided the investigators use of the
NRC's automated Document Control System for online, computer-assisted searches of NRC documentation since 1978, and all TMI-related documents. In addition, a professional technical librarian and atechnical information assistant handled investigators'
requests for specific information.
Freedom of Information Act. The NRC continues tofulfill its obligation to make available records in its
possession to interested members of the general public
who request them under the Freedom of Information
Act (FOIA). By making identifiable records available
in all cases where the requested information is not exempt from production, the NRC is contributing to fulland fair debate of public issues.
Some categories of records deemed exempt by thestatute consist of documents properly classified under
Executive Order 12065 (national security matters,
trade secrets and commercial or financial
information); some types of investigatory files; andcertain interagency or intraagency memoranda of apre-decisional nature.
The NRC continues to place material released in
response to FOIA requests in the Headquarters PublicDocument Room, where the public will have fu1l access. Additionally, documents released under the
FOIA which pertain to a particular licensed facility or
one under licensing review are furnished to the NRC
Local Public Document Room serving that facility.
The number of FOIA requests received during fiscal
year 1979 rose by 70 percent Over 1978 to 503. Thiswas due both to the public interest generated by theaccident at Three Mile Island and growing publicawareness and concern regarding various nuclear
pOwer issues. Some 16,800 staff man-hours weredevoted to supplying requesters with information requested under the FOIA, and more than 80,000 pagesof documents were released.
The Privacy Act of 1974. This law, which became
effective in 1975, provides that individuals have the247right to determine the existence of agency records concerning themselves, to seek access to them and to correct any errors that may exist. Agencies are obligatedto keep timely, accurate and complete records foragency purposes, and to advise individuals from whom
information is solicited how that information is to be
used. During fiscal year 1979, the NRC received 40
Privacy Act requests, compared with 37 received infiscal year 1978. As was the case last year, most ofthese requests came from agency employees seeking access to personnel security records about themselves.

INVOLVING THE PUBLICThe Commission took additional steps during 1979
to facilitate more meaningful and practical involvement of the public in regulatory affairs, both informally and formally.

Informal Participation
While opportunities for formal public participationin nuclear regulatory proc~edings have been provided
from the beginning and expanded periodically, the
NRC continues to seek practical means of involving
the public informally in its deliberations. In 1979, the
NRC staff sponsored conferences, workshops andregional meetings on public issues. Some of these activities were:• A meeting on the effects of low-level radiation,
held at the NRC's office in Silver Spring, Md.
Ganuary 1979).
• A workshop to discuss a proposed rule on alternate sites under the provisions of the National Environmental Policy Act, held at the Mitre Corporation in Virginia (March 1979).
• Workshops to review NRC policy on decommissioning, held in Columbia, S. C., and Seattle,
Wash. (September 1979).

The TMI-2 Investigation Center (above left) at the NRC headquarters in Washington, D.C., housed all pre- and post-accident
documentation conceming the plant. At right, a technical inform a-

In addition, the NRC conducted more than 20
public meetings from late September throughDecember 1979 on emergency response plans fornuclear power plants in operation or expected to beready for operation in the near future. The NRC sent
emergency preparedness teams to meet in theneighborhood of each plant with representatives of thelicensee and State and local officials, after notificationin the local press. An additional meeting was held toinvite public comment and questions on emergency
preparedness, lasting until everyone had an opportunity to be heard. Also, in January 1980 a series offour regional meetings to discuss proposed new
emergency response regulations were scheduled to beheld in New York, San Francisco, Atlanta, andChicago.
In addition to engaging in cooperative efforts with abroad spectrum of national and regional bodies ofState and local representatives (see Chapter 8, "State
Programs"), the NRC staff met with State legislators
numerous times during 1979 to discuss the agency's
programs on nuclear waste management, decommissioning of nuclear facilities, and radiological emergency response planning. In connection with the large
volume of legislation on nuclear power before Statelegislatures during the year, the staff provided comments when requested and presented testimony before
legislative committees on several occasions.
The event that highlighted the NRC's public affairs
in 1979, however, was the accident at Three Mile
Island. The NRC had assumed responsibility for informing1the public on the condition of the plant. Threedays after the accident, press centers were set up in the
NRC's office in Bethesda, Md., and in Middletown,
Pa. By April 1, seven NRC public affairs officers were
stationed at Middletown, where it had been decided toconcentrate all news media activities. The following
day, more than 300 national and international news

tion assistant uses the automated Document Control System tosearch out information needed by an investigator.

248

public meeting in Cedar Rapids, la., on November 14,
1978. In connection with an application for a construction permit for the New Haven Nuclear Station, apublic meeting was held in New Haven, N. Y., on environmental aspects on December 13, 1978. A public
meeting was held in Burlington, Kan., on May 15,
1979, regarding the concrete strength of the containment base mat of the Wolf Creek nuclear plant underconstruction.
Public attendance at individual sessions ranged
from four to over 200, with press and TV coverage
fairly extensive at several of the meetings. The
meetings appear to have a positive effect in permitting
the public to judge for itself the effectiveness of
nuclear regulation. The NRC intends to hold this typeof open forum whenever there is a significant technicalissue with considerable public interest or a nuclear
power plant is being considered in a vicinity for the
first time. Ways to improve anticipation of public interest continue to be explored by the staff.
Shortly after the TMI accident, press centers were set up in Mid·
dletown, Pa., and at the NRC's Bethesda, Md. office. In the photo
above, Harold Denton, Director of the Office of Nuclear Reactor
Regulation, addresses journalists in Middletown. At his left is
Joseph Fouchard, Director of the Office of Public Affairs.

media correspondents were gathered at the press
center for daily briefings on the status of the plant.
Staff Reviews Opened to Public. The use of informal
meetings has significantly expanded the opportunities
for public observance of and participation in the early
non-adjudicatory stages of the licensing process. An
NRC staff policy, which became effective in June
1978, provides that all meetings conducted by the staff
as part of the review of any proposed licensing action
are open to the general public as observers, including
parties to a hearing and petitioners for leave to intervene. In addition, selected meetings are planned
specifically to provide information to the public andare held near the proposed facility site. Suitable ad·vance arrangements and notification are made, andpublic participation (in the form of questions andcomments) is sought during these meetings.
The following are cases where such meetings were
held in fiscal year 1979 concerning nuclear power
plants. With regard to an application for a construction permit for Palo Verde Units 4 and 5, public
meetings were held in Phoenix, Ariz., on environmental matters (October 12 and 13, 1978) and on safety
matters (October 17 and 19, 1978). On full power
operation of Fort St. Vrain, open meetings were heldin Denver, Colo., on November 3 and 4, 1978. A program for repair of the recirculation inlet safe ends atthe Duane Arnold Energy Center was the subject of a

UGovemment in the Sunshine." During fiscal year
1979, the Commission opened two-thirds of its
meetings to public observation in compliance with theGovernment in the Sunshine Act. The statute, whichbecame effective on March 13, 1977, regulates the
conduct of meetings of collegial agencies like the ,NRC,
and makes their deliberative processes more accessible
to the public.
The Commission's regulations implementing theSunshine Act (10 CFR Part 9, Subpart C) specify procedures for deciding whether to close a meeting, what
records will be kept, and other administrative details.
The law requires the Commission to open all of its
meetings to public attendance unless one or more of 10
exemptions applies. The exemptions are designed topermit closed discussion of specified matters.
However, transcripts or recordings must be made ofmost closed meetings and are released to the public
when appropriate. In closing one-third of its meetingsin fiscal year 1979, the Commission primarily citedfour of the 10 exemptions: 1 (classified information), 2(internal rules and practices), 6 (personnel) and 10(adjudicatory/litigation). The regulations also specify
that advance notices of meetings be published in theFederal Register, placed in the Public Document
Room, and mailed directly to individuals andorganizations on request.
The Commission firmly supports the principles ofopen government enunciated in the Sunshine Act andhas voluntarily chosen to go beyond the literal requirements of the Act to adopt policies that advance its purposes. For exampfe, staff papers and documentation
pertaining to the proposed issuance of export and import licenses are made available in the PDR; someCommissioners' correspondence is placed in the PDR;

249

staff papers discussed in public Commission meetings
are placed in the PDR, together with handouts andvisual material presented at the meeting; radio
coverage, television coverage and tape recordings ofCommission and licensing board meetings are permitted; and, in cases of general interest, the public has
been permitted to attend Commission adjudicatory
sessions that could have been closed under Exemption
10 of the Sunshine Act. In addition, the Commission
has a continuing program for reviewing transcripts ofclosed meetings so they may be released to the public.
Since the enactment of the Sunshine Act, the Commission has released 325 transcripts of closed meetings, including adjudicatory minutes.

Formal Public Participation
NRC regulations provide for formal participation
by members of the public as parties in rulemaking,licensing and other proceedings. Opportunities forhearings are indicated in the accompanying table.
Commission regulations require that a public hearing on each application for a major nuclear facility
construction permit be conducted by an Atomic Safety
and Licensing Board (see Chapter 13). The hearing isannounced well in advance in the Federal Register andposted in a public document room near the proposedconstruction site, together with a copy of the application. Local newspapers also carry notice of the hearing. Interested persons or groups are invited to participate in the hearing by: (1) submitting a written
statement at the hearing; (2) making an oral presentation at the hearing; or (3) petitioning the licensing
board for the right to become an "intervenor" in theproceeding with full participatory rights, including
cross-examination of other participants. Intervenors
participate fully in prehearing conferences with otherinterested parties for the exchange of data and identification of issues in contention.
If the licensing board disallows a petition, the petitioner may appeal to the Atomic Safety and Licensing
Appeal Board (see Chapter 13). In some instances, theCommission may rule on a petition. Ultimately, a petitioner may seek a ruling in the appropriate Federal
Court of Appeals and the Supreme Court of the UnitedStates.
The same rights and procedures for public participation apply to hearings on applications for
operating licenses, with the difference that such hearings are not mandatory and need not take place unlessrequested by one or more interested parties.
To facilitate public participation, hearings of thelicensing boards, with rare exceptions, are held incommunities near each proposed facility site. To
enhance public participation in the review and hearing process for facility license applications, and to improve coordination with States, counties, andmunicipalities, the NRC's basic rules of practice (10CFR Part 2) contain the following key elements~• Interested persons may make limited appearances
at prehearing conferences.
• Interested cities, counties and local government
agencies may participate in licensing proceedings
without taking a position on the issues, a privilege
also accorded to States.
• Interested States, counties, cities and/ or agencies
thereof may file proposed findings of fact andconclusions of law, exceptions to initial decisions,and petitions for Commission review.
• Procedures have been established for amicus participation in appeals before licensing boards or
the Commission.
• Motions for summary disposition are no longer
limited to initial licensing proceedings.
• Licensing boards have authority to consolidate
two or more proceedings for hearing.
• Joint hearings with States or other Federal agencies are authorized on matters of concurrent
jurisdiction; however, NRC rules of practice may
not be waived, and the action must be conduciveto the proper dispatch of Commission businessand the ends of justice.

NRC's Director of Nuclear Reactor Regulation, Harold Denton,
responds to questions from the media in Middletown, Pa., a few
days following the onset of the TMI accident.

250

ENHANCING INTERNALCOMMUNICATION

The following list shows the date, committee, and subject of each hearing.

Since the NRC's inception, its Chairmen have supported an "open door" policy for the consideration of
the views of all employees which extends up throughthe management chain to the Commissioners' offices.
Safety matters may also be discussed with the independent Advisory Committee on Reactor Safeguards.
However, in a July 1978 memorandum to all
employees, Chairman Hendrie asked for comments
and suggestions which might help "to make the 'open
door' policy more of a reality both in concept and inpractice."
Included in the memorandum were the results of asurvey of policies and procedures used or considered
by a number of Federal agencies, business corporations, professional societies and other privateorganizations for bringing differing professional views
to the attention of management, and for appropriate
management response (NUREG-0500).
The survey described concepts that NRC planned touse in developing formal procedures for making
known to management employees' opinions on any
substantive matter within the agency's purview thatdiffer from an existing policy or' a proposed staff position on the matter. It identified and discussed procedural steps that could provide for: making employee
differences known to management, management
response, alternatives if the employee is dissatisfied
with the response, follow-up on resolution of the issue,
and follow-up to ensure that the employee is not subjected to retaliatory actions. In addition, the survey
described criteria that could be used to judge the effectiveness-and perhaps the acceptability-of any
mechanism designed to handle differing professional
opinions.
Comments were solicited from both NRC employees
and the public for consideration in developing anagency-wide plan (NUREG-0567). In late 1979, a supplemental statement to this revised plan on differing
professional opinions was sent to all NRC employees
for comment in early 1980.

10/ 3/78-House Committee on Foreign Affairs, Subcommittee on International Economic
Policy and Trade (Nuclear Fuel Transfer
for Reprocessing)
1I26179-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and theEnvironment (Spent Fuel)2/ 5/79-Senate Committee on Environment andPublic Works, Subcommittee on Nuclear
Regulation (NRC Authorization)
2/22/79-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and theEnvironment (NRC Authorization)
2/26/79-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and Environment (Reactor Safety Study)
2/27!79-House Committee on Interstate andForeign Commerce, Subcommittee on
Energy and Power (NRC Authorization)
3/ 5/79-House Committee on Appropriations, Sub.
committee on Energy and Water Development (NRC Appropriations)
3/ 13/79-Senate Committee on Appropriations, Subcommittee on Energy and Water Development (NRC Appropriations)
3/14!79-Senate Committee on Governmental Affairs, Subcommittee on Energy, Nuclear
Proliferation and Federal Services(Disposal and Storage of Spent Nuclear
Fuel and Waste)
3/16/79-Senate Committee on Environment andPublic Works, Subcommittee on Nuclear
Regulation (NRC Shutdown of Five
Nuclear Reactors)
3/19/79-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and theEnvironment (NRC Shutdown of Five
Nuclear Reactors)
3/21179-Committee on Appropriations, Subcommittee on Energy and Water Development(NRC Shutdown of Five Nuclear Reactors)
3/27/79-Senate Committee on Environment andPublic Works, Subcommittee on Nuclear
Regulation (NRC Authorization)
3/29/79-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and theEnvironment (Three Mile Island Accident)
41 4/79-Senate Committee on Labor and Human
Resources, Subcommittee on Health andScientific Research (Three Mile Island Accident)
4/10/79-Senate Committee on Environment andPublic Works, Subcommittee on Nuclear
Regulation (Three Mile Island Accident)

CONGRESSIONAL OVERSIGHT
The number of hearings by the several Congressional committees exercising jurisdiction over NRC activities continued to increase in 1979. During the fiscal
year, NRC witnesses testified a total of 42 times before
18 committees or subcommittees on such subjects as
the Three Mile Island accident, the health effects ofradiation, the shutdown of five facilities due to seismic
problems, emergency preparedness, safeguards, andwaste management. The NRC testified at an additional13 hearings in October through December 1979.

251

OPPORTUNITIES FOR FORMAL PUBLIC HEARINGS IN NRC PROCEEDINGSType oj
Proceeding

Opportunity
jor Hearing

Purpose oj
Hearing

Criteria jor
Granting Hearing

Unit Deciding
To Hold Hearing

RULE MAKING
Proceeding

Prior to issuance of
final rule.

To determine whether
a proposed rule should
be adopted.

At the discretion of
the Commission.

Commission (which
may decide to holdinformal of "hybrid"hearing).

MANUFACTURING LICENSE
Proceeding·

Mandatory hearing
prior to issuance of
manuf acturing
license.

To determine whether
a license authorizing
the manufacture of a
production or utilization facility of a particular design should
be issued.

Mandatory hearing on
safety and environBoard.

Mandatory hearing
before Licensing
Board.

CONSTRUCTIONPERMIT
Proceeding·

Mandatory hearing
prior to issuance ofconstruction permit.

To determine whether
a particular production or utilization
facility should be constructed at a particularsite and, where indicated, to resolve
adverse antitrust
matters.

Mandatory hearing on
safety and environmental issues; on antitrust matters, uponrequest by interested
persons or Attorney
General or at discretion of Commission.

Mandatory hearing
before Licensing
Board.

OPERATING
LICENSE
Proceeding*Prior to issuance of
operating license.

To determine whether
a particular production or utilization
facility should be permitted to operate;
antitrust review where
significant changes
have occurred sinceprevious antitrust
review.

Request by any person
whose interest may beaffected by proceeding
who raises genuine issue of material fact,
and at discretion ofCommission; in addition, in the case of
antitrust review, there
must be determinationby the Commissionthat significant
changes have occurred.

Commission, Appeal
Board or Licensing
Board, as appropriate.

MATERIALS
LICENSE
Proceeding

Either prior to or afterissuance of materials
license.

To determine whether
a particular materials
license should be issued or remain in
effect.

Request by any person
whose interest may beaffected by proceeding and at discretionof Commission.

Commission, Appeal
Board, Licensing
Board or Administrative Law Judge, as
appropriate.

SHOW CAUSE
Proceeding (to modify,
suspend or revoke alicense pr for otherappropriate action).

Prior to issuance of
final Commission
Order.

To determine appropriate action to betaken.

Upon demand by person cited in Show
Cause Order or by request of other persons
whose interest may beaffected, upon making
requisite factual
showing.

Commission.

.. An opportunity for hearing is also provided prior to issuance of amendments to manufacturing licenses, construction permits and operating
licenses which involve significant hazards considerations. If there are no significant hazards considerations, opportunity for hearing may beprovided after such amendments are issued.

252

4/30/79-Senate Committee on Environment andPublic Works, Subcommittee on Nuclear
Regulation (Three Mile Island Accident)
51 8/79-Senate Committee on Governmental Affairs, Subcommittee on Energy, Nuclear
Proliferation, and Federal Services (Libassi
Report on Low-Level Radiation)
51 9/79-Senate Committee on Government Affairs,
Subcommittee on Energy, Nuclear Proliferation, and Federal Services (Federal
and State Radiation Monitoring at TMI.and Issues Relating to the Siting of Nuclear
Reactors)
5/10179-Senate Committee on Energy and Natural
Resources (Nuclear Waste Policy
Act-S.685)
5/14179-Committee on Government Operations,
Subcommittee on Environment, Energy
and Natural Resources (Emergency
Preparedness at NRC-Licensed Nuclear
Facilities)
5/16/79-House Committee on Armed Services, Subcommittee on Military Installations andFacilities (Civil Defense Preparedness for
Three Mile Island)
5/21179-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and theEnvironment (TMI and Nuclear Reactor
Safety)5/22179-House Committee on Science andTechnology, Subcommittee on Energy
Research and Production (Safety Designs of
Nuclear Power Plants)
5/23179-House Committee on Science andTechnology, Subcommittee on Energy
Research and Production (Safety Designs of
Nuclear Power Plants)
5/31179-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and theEnvironment (NFS West Valley Issues)
61 2179-House Committee on Science andTechnology, Subcommittee on Natural
Resources and Environment (Health Effects of TMI)
61 4179-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and theEnvironment (Nuclear Regulation)

At the outset of an April 10, 1979 hearing on the TMI accident,
Senator Gary Hart, Chairman of the Subcommittee on Nuclear
Regulation of the Committee on Environmental and Public Works,

swears in the five NRC Commissioners. NRC witnesses appeared.
at more than 40 hearings during fiscal year 1979.

253

61 6/79-House Committee on Post Office and Civil
Service, Subcommittee on Investigations(Reactor Construction Si tes- Manpower
Utilization)
6/11179-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and Environment (Domestic Nuclear Industry
Security)
6/13/79-House Committee on Science andTechnology, Subcommittee on Natural
Resources and Environment, and Subcommittee on Energy Research and Production(Low-level Ionizing Radiation)
6/14/79-Senate Committee on Environment andPublic Works (Confirmation of Victor
Gilinsky)
6/22/79-Senate Committee on Foreign Relations,
Subcommittee on Arms Control, Oceans
and International Operations and Environ.
, ment (IAEA: Safeguards)
6/27/79-House Committee on Interstate andForeign Commerce, Subcommittee on
Energy Power (Storage and Disposal of
Spent Nuclear Fuel)
6/28/79-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and theEnvironment (Public and State Involvement in Waste Management)
71 9/79-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and theEnvironment (Price-Anderson Act andLiability for Nuclear Incidents)
7/19/79-Senate Committee on Commerce, Scienceand Transportation, Subcommittee on
Science, Technology and Space (Safety in
the Transportation of Radioactive
Shipments).
7/19/79-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and theEnvironment (State Regulations of Nuclear
Activities)
7/26/79-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and theEnvironment (Nuclear Proliferation)
8/10179-House Committee on Interior and Insular
Affairs, Subcommittee Oversight and Investigations (Waste Isolation Pilot Project)
9/11179-Senate Committee on Environment andPublic Works Subcommittee on Nuclear
Regulation (Nuclear Waste Management)
9/19/79-House Committee on Science andTechnology, Subcommittee on Energy
Research and Production (Improved Safetyof Nuclear Power Plants)
101 2/79-Senate Committee on Environment andPublic Works, Subcommittee on Nuclear
Regulation (TMI Investigation)

101 3/79-Senate Committee on Environment andPublic Works, Subcommittee on Nuclear
Regulation (TMI Investigation)
101 5/79-Senate Committee on Foreign Relations(U. S. - Australian Agreement for Cooperation)
10/11179-House Committee on Foreign Affairs, Subcommittee on International Economic
Policy and Trade (U.S.-Australian Agreement for Cooperation)
10/22/79-House Committee on Interior and Insular
Affairs, Subcommittee on Energy and theEnvironment (Uranium Mill Tailings
Disposal at Church Rock)III 1I79-House Committee on Government Operations, Subcommittee on Environment,
Energy and Natural Resources (Emergency
Preparedness at TMI)
III 5/79-House Committee on Interstate andForeign Commerce, Subcommittee on
Energy and Power (Kemeny Commission
Report-TMI)
III 7/79-House Committee on Science andTechnology, Subcommittee on Energy
Research and Production (Low-Level
Radioactive Waste Disposal)
III 8/79-Senate Committee on Environment andPublic Works, Subcommittee on Nuclear
Regulation (Kemeny Report)
III 9/79-Senate Committee on Environment andPublic Works, Subcommittee on Nuclear
Regulation (Kemeny Report)
11114/79-House Committee on Science andTechnology, Subcommittee on Energy
Research and Production (TMI-Kemeny
Report)11127/79-House Committee on Government Operations, Subcommittee on Environment,
Energy and Natural Resources (Marble
Hill)

12/11179-Senate Committee on Environment andPublic Works, Subcommittee on Nuclear
Regulation (Waste Management)

Reports to Congress
The NRC .must keep committees having jurisdiction
over its functions under rules of the Senate and theHouse "fully and currently informed" regarding theCommission's activities. Information on significant
developments is forwarded routinely to the appropriate committees, and special reports are issued in
response to inquiries by committees and individual
members of Congress.
Periodic reports to Congress or Congressional committees are required by law on the following matters:

254• NRC Annual Report to the President, for histransmittal to the Congress on a fiscal year basis.
• Abnormal occurrences in regulated nuclear activities (quarterly).
• Indemnity activities under the Price-Anderson
Act (annual; now being incorporated in theoverall Annual Report).
• Administration of the Freedom of Information
Act (annual).
• Implementation of the Government in the Sunshine Act (annual).
• Printing plant report (annual).
• Annual plant inventory (annual).
• Major organizational components and numbers of
employees (annual) .
• Steps to meet provisions of Equal Opportunity
Act (quarterly).
• Progress on resolving generic safety i~ssues related
to nuclear power plants (annual; being incorporated in the NRC Annual Report).
• Updating of long-term research plan for projects
to develop new or improved safety systems fornuclear power plants (annual).
• Commission's views and recommendations on
U.S. policies and actions to prevent proliferation(annual).
• ACRS report concerning nuclear reactor safety
research program (annual).
In addition, the fiscal year 1979 NRC Authorization
Act contained provisions affecting the NRC's activities
and authority, and required new reports to Congress,
both through new mandates and through amendments
to the Atomic Energy Act of 1954 and the Energy
Reorganization Act of 1974. The principal new reporting requirements are as follows:• Agency capabilities and research regarding thehealth effects of low-level radiation (April 1979) ,and options for Federal research in this area
(September 1979).
• Status of domestic safeguards matters during
previous fiscal year (annual; incorporated inoverall Annual Report).
• Fuel cycle systems evaluation (semi-annual; annually in 1981 and 1982).
• Radioactive waste storage or disposal activities(March 1979).

• Agency use of contractors, consultants, and national laboratories (annual).
• Review of selection and training for members ofAtomic Safety and Licensing Boards (1979).
GAO Reports. A number of other Congressional
reports are issued as the result of studies by theGeneral Accounting Office under its broad authority
to assist Congress, its committees, and individual
members in carrying out their legislative and oversight
responsibilities.
An agency which is the subject of GAO reports
recommending corrective actions is required by law toreport within 60 days to the Government Operations
Committees of the House and Senate on steps taken orplanned to implement the recommendations. During
fiscal year 1979, the GAO issued seven reports covering various aspects of NRC activities. Four more were
issued in October through December 1979. NRC
responses to GAO recommendations are available inthe main NRC Public Document Room in Washington, D.C. GAO reports issued during the year are:121 18/78-"Nuclear Diversion in the U.S.? 13 Years ofContradiction and Confusion."
1123/79-" Automated Systems Security-Federal
Agencies Should Strengthen Safeguards
Over Personal and Other Sensitive Data."
1I26179-"Reporting Unscheduled Events at Commercial Nuclear Facilities: Opportunities
to Improve Nuclear Regulatory Commission Oversight."
2/16/79- "Higher Penalties Could Defer Violationsof Nuclear Regulations."
31 8179-Letter report on NRC's use of DOE
Laboratories and of outside contractors
and consultants.
3/30/79-"Areas Around Nuclear Facilities Should
Be Better Prepared for Radiological
Emergencies ...
51 7179-"Federal Actions Are Needed to Improve
Safety and Security of Nuclear Materials
Transportation. "101 2179- "Emergency Preparedness Around the Rancho Seco Nuclear Powerplant: A Case
Study."
10/10/79- "Nuclear Construction Times for the Second and Subsequent Plants at a MultiPlant Site are Overstated."
11115179-"Placing Resident Inspectors at Nuclear
Power Plants: Is It Working?"
121 4179-"Radiation Control Programs Provide
Limited Protection."

13
Proooedings andLitigation
Structural pilings were the issue in the latest
hearings on the Bailey Generating Stationin Indiana.

The following are accounts of adjudicatory activityof the NRC during fiscal year 1979, with highlights
through December, covering specifically activities of
the Atomic Safety and Licensing Boards, the Atomic
Safety and Licensing Appeal Boards, and the Commission. In addition, brief reviews are presented ofFederal court actions in which the NRC was a party or
had an interest.
The most significant developments affecting the adjudicatory phase of the licensing process were delays inseveral proceedings before licensing boards pending
the NRC stafrs evaluation and the boards' review of
the Three Mile Island nuclear power plant accident,
and the Commission's announcement of October 5that no licensing board decisions authorizing issuanceof a construction permit, limited work authorization
or operating license would be issued except after further order of the Commission itself. The Commission
action was part of an interim policy statement settingout procedures to be followed while the Commission
considers a range of options dealing with the extent towhich its regulatory structure should be modified as aresult of the TMI accident. (See Chapters 1 and 2.)
Study of the "Immediate Effectiveness Rule." In its
January 1978 Seabrook decision, the Commission
spoke of the need to reassess the wisdom of its immediate effectiveness rule (10 CFR 2.764) which permits construction or operation to begin at a nuclear
plant immediately after a favorable licensing board
decision even though an appeal from that decision
may be pending before an appeal board or before theCommission. In January 1979 the Commission
established a ten-member advisory committee, chaired·by Professor Gary Milhollin of the University of
Wisconsin School of Law, to study the immediate effectiveness rule and to report on alternatives. The'
committee held more than a dozen public meetingsand obtained public views through a workshop and a

Federal Register solicitation of comments. In
December 1979 the advisory committee issued its final
report and briefed the Commission on its findings andrecommendations. The Commission accepted therecommendations of the committee and directed the
Office of the General Counsel to work with Professor
Milhollin to prepare a Notice of Proposed Rulemaking
to seek public comment on the three alternatives to therule which were proposed by the committee.
Study of the NRC Appellate System. In fall of 1978,
the Chairman requested that the Office of the GeneralCounsel prepare a study of the Commission's appellate
adjudicatory system in order to analyze the merits ofproposals to abolish the Atomic Safety and Licensing
Appeal Panel or to combine the functions of that withthose of the Atomic Safety and Licensing Board Panel.
The General Counsel's study was completed in
December 1979, and the Commission was briefed on
the results of the study in January 1980. The study examined the History of the Appeal Panel, the role of the
Appeal Panel in the Commission's overall adjudicatory
system, the Panel's workload, and the practices ofother Federal agencies with similar responsibilities. A
number of alternatives to the present adjudicatory
system were identified and analyzed in the study. The
General Counsel's conclusion was that the current adjudicatory system, including the Appeal Panel, should
be retained but with several modifications designed toincrease the opportunity for early Commission-level
involvement in proceedings involving significant newissues of law or policy.

ATOMIC SAFETY AND LICENSING
BOARDS
Public participation in the licensing process is apparent in proceedings conducted by Atomic Safety andLicensing Boards, for it is here that individuals may

256

voice their interests about a particular licensing issue
before an independent tribunal that will consider theirconcerns before rendering a decision.
The Atomic Energy Act of 1954 requires that apublic hearing be held on every application for a construction permit for a nuclear power plant or related
facility. An independent Atomic Safety and Licensing
Board conducts this hearing. This board issues a decision on the application (known as an "Initial
Decision"), which, subject to the NRC's review andappellate procedures, may become the final NRC decision. The hearing announcement, which invites public
participation, is published shortly after receipt of aconstruction permit application so that interested parties may be aware of the proceeding at an early stage.
The announcement is given to appropriate State andlocal agencies, as well as other interested groups.
Commencement of the hearing itself must await thecompletion of the NRC staff's safety or environmental
review.
The Atomic Energy Act requires that a second opportunity for hearing be provided before a license may
be issued to operate a faCility. A similar opportunity isprovided before certain license amendments may be
issued. Public participation is also invited in proceedings instituted by the NRC staff.
The Atomic Energy Act also requires that, prior to
the issuance of a construction permit for a nuclear
power plant or related facility, a determination bemade by NRC as to whether the activities licensed by it
would create or maintain a situation inconsistent withthe antitrust laws. While the procedures for thisreview are more complex than those for other reviews,
an opportunity to request a hearing before a licensing
board is provided to those whose interests may be af.
fected.
Each of the licensing boards consists of threemembers drawn from the membership of the Atomic
Safety and Licensing Board Panel-a b9dy of legal,
technical, environmental, and other experts appointed
by the Commission. As of September 30, 1979, thePanel included 13 full-time and 40 part-timemembers. Of these 53 members, 18 are lawyers, 16 environmental scientists, 10 engineers, 7 physicists, 1
economist and 1 chemist. (See Appendix 2 for names ofmembers.) The Commission appoints members to the
Panel based upon recognized experience, achievement, and independence in the appointee's field. In
assigning individuals to a licensing board, consideration is given to the kinds of issues involved in the proceeding before that board. Generally, boards consistof a lawyer-chairman, a nuclear engineer or reactor
physicist, and an environmental scientist. However,antitrust problems are heard and decided by a boardof three antitrust experts.
Aside from the hearing on antitrust matters, a hearing on a particular application may be divided into
two phases-one concerning the health and safety,

and common defense and security aspects of the application, as required by the Atomic Energy Act; andthe other concerned with the environmental considerations required by the National Environmental
Policy Act. Separate Initial Decisions covering thesematters may be issued.
During the period covered by this report, boards
were increasingly occupied by operating license andlicense amendment proceedings and less occupied byconstruction permit proceedings. Only two construction permit decisions, authorizing four units (Yellow
Creek 1 and 2 and Jamesport 1 and 2), were issued. At
the same time, petitions to intervene and requests forhearing were received in 12 of 31 operating license andlicense amendment cases which were noticed. Of
these, 10 resulted in hearings being ordered, one wasdenied, and one was withdrawn. One operating
license decision was issued that authorized the operation of the McGuire station's two units. The authorization was, however, stayed pending receipt and review
by the board of a supplement to the staff's Safety
Evaluation Report.
The Diablo Canyon partial initial decision covering
seismic, security, and potential aircraft crash issues,
fell into this category because it withheld ruling on theoperating license pending receipt of the staff's report.
After the Three Mile Island accident, construction
permit and operating license proceedings nearing decisions were held up awaiting the staff's evaluation andthe board's review of that accident. Boards, however,approved two operating license amendments authorizing expansion of reactor spent fuel pools. Additionally,four proceedings were pending before boards whereapplicants sought to obtain an early hearing and decision on certain site-related matters. Two of these werecommenced during the report period.
Important issues heard and decided by boards during the report period included the following:• In Troian, it was discovered during the course ofcertain modifications to the plant that some structural features did not meet seismic requirements.
Pursuant to a request from members of the
public, a hearing was held before a board todetermine whether interim operation of the plant
could be permitted pending approval and implementation of the necessary modifications.
After hearing all the evidence, the board permitted operation to resume on an interim basis subject to certain conditions. The board has scheduled a hearing to review the modifications proposed to meet seismic requirements for April
1980.
• In the proceeding on Offshore Power Systems' application to manufacture floating nuclear power
plants, the board ruled that an environmental impact statement covering the entire manufactUring
program for the plants was not required. Under

257

The Pacific Gas and Electric Co.'s Diablo Canyon Nuclear Power
Plant, is situated on the Pacific Coastline near San Luis Obispo,
California. It was the subject of public hearings early in 1979 on
plant seismic capabilities in light of its location less than five miles

from an offshore geologic fault. At year's end, Unit 1 had been
completed and Unit 2 was about 98 % complete, with completionof modifications based on the seismic reevaluation expected by
mid-1980.

the Supreme Court's ruling in Kleppe v. Sierra
Club, the board held that the environmental impact statement need cover only the specific proposal to manufacture eight plants, noting theCommission staffs statement that a further impact statement would be prepared on any proposal to manufacture additional plants.
• In McGuire, an organization petitioned to intervene as a representative of its members in alicense amendment proceeding. In its petition,
the organization challenged the requirement that
it identify certain of its members who met thelegal requirements for participation in the proceeding on the ground that to do so would violate
certain of these individuals' rights. The board
reaffirmed the requirement that at least one suchmember be identified.

• In another intervention ruling, involving Unit 2of the Washington Public Power Supply System, aboard was presented with a petition by anorganization which did not meet the legal requirements for intervention because it lacked
members residing in the vicinity of the plant concerned. The board ruled that this defect could not
be cured by the organization's acquisition of suchmembers well after the time for filing of petitions
had expired.
• In the Fermi 2 proceeding, the board was facedwith the applicant's request for discovery of thefacts relating to the legal basis for a petitioner's
right to participate in the proceeding. This request was made prior to the board's ruling on the
petition. The board ruled that such discovery was
not authorized by the rules.

258

An NRC Atomic Safety and Licensing Board conducted public
hearings in California during Fiscal Year 1979 on the seismic
capabilities of the Diablo Canyon Nuclear Power Plant. The hearings followed extensive review by the NRC staff and Advisory
Committee on Reactor Safeguards of considerations associated with
the Hosgri Fault, located some 3.5 miles offshore from the plant
ASLB members Elizabeth Bowers (presiding), William Martin(left), and Glenn Bright are shown.

Other rulings of interest include ones which passedon the qualifications of an expert witness (Diablo Canyon); which held that fuel could be delivered to theplant site prior to approval of the operating license(Zimmer); and which refused to prevent processing ofa~ e~rl);' site review application, pursuant to the CommisslOn s rules (Fulton). Boards also issued five orders
ruling on motions for summary disposition in which
they determined whether issues which had been raisedin proceedings could be decided without the necessityof an evidentiary hearing. In these orders, the boards
were able to decide some issues without a hearing,
while setting others down for evidentiary presentations because the nature of the facts surrounding theseissues required further elucidation.
At year-end, there were five action antitrust proc~edings pending before boards, four in the prehearing
dIscovery phase and one the subject of settlementnegotiations among the parties.

ATOMIC SAFETY AND
APPEAL BOARDS

LICENSING

Atomic Safety and Licensing Appeal Boards, consisting of three members each, perform the Commis-

sion's review functions in facility licensing proceedings
and in such others as the Commission may specify.
Board membership for each proceeding is selectedfrom among the members of the Atomic Safety andLicensing Appeal Panel by the Chairman of the Panel.
(See Appendix 2 for membership of the Panel.)
Appeal boards entertain appeals from Initial Decisions of licensing boards and certain licensing board
orders pertaining· to petitions by members of the
public seeking to intervene in NRC licensing proceedings. They also review Initial Decisions on theirown initiative and sometimes consider questions on
rulings referred by a licensing board while the proceeding before it is still in progress. Appeal boards occasionally conduct evidentiary hearings as part of theirappellate review functions or as directed by the Commission. The appeal board is the highest level within
the Nuclear Regulatory Commission at which a party
may seek administrative review as a matter of right.
Parties are permitted, however, to seek discretionary
Commission review of certain appeal board rulings.
The Commission also may itself decide to review anappeal board action. If the Commission does notreview a decision, the decision of the appeal board
becomes the final order of the Nuclear Regulatory
Commission, subject to review in a Federal court ofappeals.
During fiscal year 1979, the appeal boards issued 64
decisions and orders, which were included in the
Nuclear Regulatory Commisson Issuances, thepublication containing the adjudicatory issuances of
the NRC. Numerous other unpublished orders,
generally procedural in nature, were also issued by theappeal boards in the course of conducting the proceedings before them.
The past year presented the appeal boards withmany issues concerning environmental acceptability
and compatibility with public health and safety. In
the environmental area, developments in the muchpublicized Seabrook (New Hampshire) proceeding
were of particular significance. The appeal board held
hearings near the Seabrook site on alternate locations
for the Seabrook facility. Subsequently, the Court ofAppeals for the First Circuit in New England Coalition on Nuclear Pollution v. NRC (582 F.2d 87), issued
its decision upholding an earlier Commission rulingthat in comparing construction costs at the Seabrooksite with those at alternate sites, actual costs to complete the facility should be used. The Court's decision
ended the need for further appeal board review of thealternate site issue.
Significant environmental issues were also raised inother proceedings. The Hartsville (Tennessee) proceedings, involving a facility of the Tennessee Valley
Authority, dealt with the effects of a discharge diffuserin the Cumberland River on an endangered species ofmussels (Lampsilis orbiculata). The diffuser will carrywaste and service water from the plant to the river
where the mussels live. Following agreement among..

==============~~====~====~==================~=259

all parties on a monitoring plan to protect the mussels
during construction of the diffuser, the appeal board
issued a decision allowing its construction. Phipps
Bend (Tennessee), involving another TVA facility,.
concerned the respective responsibilities of TV A andthe NRC for satisfying requirements of the National
Environmental Policy Act (NEP A). The appeal board
affirmed a licensing board ruling that, as the licensing
agency, the NRC has the jurisdiction to impose license
conditions designed to mitigate adverse environmental
effects resulting from the construction and operationof a TV A plant. And Yellow Creek (Tennessee),another proceeding involving a TVA facility, raised
another question concerning the respective responsibilities of two Federal agencies under the environmental laws-in this instance, the responsibilities of theEnvironmental Protection Agency and the NRC under
NEPA and the Federal Water Pollution Control Act(FWPCA). The appeal board ruled that the NRC may
not incorporate in licenses to build nuclear power
plants conditions that call for a review of the adequacyof water quality requirements previously establishedby EPA under the FWPCA.
In Marble Hill (Indiana), the issue was which of two
States, Kentucky or Indiana, was the proper State forFederal Water Pollution Control Act certification purposes. This issue arose as a result of a conflict between
the two States over the boundary line separating themalong the Ohio River, close to the plant. The appeal
board decided that the boundary line placed theplant's discharge pipe within Indiana's borders, andtherefore, the applicant's certification from that Statemet the Water Act's requirements.
The environmental and health effects of radon(Rn-222) releases produced in the mining and millingof uranium continued to require the attention of theappeal boards in the Peach Bottom (Pennsylvania) anda number of other proceedings. The appeal boards
jointly handed down orders refining the procedural
framework established earlier for resolving the issue
without having to hold separate repetitive trials ineach pending proceeding. This effort led to a consolidation of several of the cases and disposition,
without hearing, of a number of items in dispute.
Three members of the Appeal Panel were selected by
the Panel membership to take evidence on the remaining, unresolved items. The need for consideration ofthis matter stemmed from an April 1978 issuance inwhich the Commission decided that the effects ofradon considered by licensing boards in each of 17
earlier proceedings should be reevaluated, because thelicensing board evaluations had been made on thebasis of a rule that the Commission later found to be inerror.
Evidentiary hearings were held by an appeal board
in a North Anna (Virginia) proceeding on two safety
questions, one of which was raised by the board in its
review of the licensing board decision authorizing an

operating license for the plant. It concerned the plant's
ability to withstand damage from missiles generated
inside or outside the plant. The second issue, raised byan intervenor, involved the settlement of the land
under parts of the site. Decision on these issues remained pending at year-end.
Another proceeding in which evidentiary hearings
were conducted by an appeal board involved the question of whether a cloud of flammable vapor from atanker accident on the Delaware River might reachthe Hope Creek (Pennsylvania) plant. The appeal
board ruled that the probability of such an occurrence
was so small that it need not be considered in thedesign of the plant.
The appeal board in Callaway (Missouri) ruled that
the NRC could suspend the construction permit for theplant for failure of the construction contractor tocooperate in NRC's investigation into the firing of aconstruction worker" allegedly for "whistle blowing"on unsafe construction practices. The appeal board
held argument on this appeal in the Moot Court Roomof the Howard University School of Law in
Washington, D.C., with students and faculty
members of the school's administrative law and environmental law classes in attendance.
During the year the appeal boards elaborated on therequirements that persons and organizations desiring
to intervene in NRC licensing proceedings must meet.
In the North Anna (Virginia) operating license amendment proceeding in which the licensee was seeking
permission to expand the capacity of the facility's
spent fuel pool, a petitioner's close proximity to thefacility was held to be enough to establish the requisite
interest for intervention. Allens Creek (Texas) dealt
with the appeal of various persons and organizations
whose intervention petitions in this reactivated construction permit proceeding had been denied by thelicensing board. In ruling on the petitions underappeal, the appeal board detailed the requirements
for an organization whose standing to intervene
is dependent upon at least one of its members. And in
Skagit (Washington), with one member dissenting, theappeal board affirmed the licensing board's denial of alate intervention petition of three Indian tribes.
Earlier, the licensing board had granted the Indiansintervention on the ground that the special relationship between the United States and Indians compelled
the grant of the petition even though it was more than
three years late. The appeal board held this ruling to
be in error and set the decision aside. Upon reconsideration, the licensing board denied the petition andthis action was affirmed by the majority of the appeal
board. This matter is now being considered by theCommission.
the past year also presented the appeal boards with
important questions about the NRC's subpoena
authority. In the Stanislaus (California) antitrust proceeding, the appeal board was concerned with260

whether a subpoena duces tecum could be issued to
one who was not a party to the proceeding. The appeal
board held that such a subpoena was authorized by theCommission's regulations and the Atomic Energy Act.
It further held that the Commission could require theparty requesting this subpoena to reimburse the subpoenaed person for its cost in reproducing the documents demanded in the subpoena, but it denied the request by the subpoenaed company for reimbursementof its search costs on the ground that those costs werereasonably incident to the conduct of its business.
Diablo Canyon (California) raised the question. whether "exceptional circumstances" existed to justify
the issuance of subpoenas, pursuant to 10 CFR 2.720,
compelling the testimony of two consultants to the Advisory Committee on Reactor Safeguards. In rendering
advice to the ACRS, these consultants had expressedviews contrary to the opinion adopted by the ACRS.
In the circumstances involved (significant intervening
seismic-related developments since construction of theplant was first approved), the appeal board concluded
that the subpoenas were appropriate.
In a civil penalty proceeding involving the Atlantic
Research Corporation, the appeal board overturned
the decision of the NRC administrative law judge
affirming the imposition of a total of $8,600 in civil
penalties against the company. The penalty had been
imposed by NRC's Director of the Office of Inspection
and Enforcement largely because of the radiation exposure suffered by a company radiographer due to theemployee's own unauthorized acts. In setting aside thecivil penalties in their entirety, the appeal board foundno evidence of any shortcoming by the licensee, and,in the circumstances involved, it perceived no discernible remedial purpose that would be served by the civil
penalties.

COMMISSION DECISIONS
Final Fuel Cycle Rule PromulgatedOn July 27, 1979, the Commission promulgated a
final rule -.vhich sets out a table of revised environmental impact values for the uranium fuel cycle ("Table
S-3"). These values are to be included in environmental reports and impact statements for individual lightwater nuclear power reactors. (See 10 CFR Part 51
and 44 Fed. Reg. 45362 of August 2, 1979.) The
revised Table S-3 replaces an interim Table S-3 promulgated for a limited period on March 14, 1977. The
final rule became effective on September 4, 1979.
This final rule is the product of an extensive
rulemaking which began May 26, 1977, and included
public hearings before a three-person hearing board
and oral presentations to the Commission itself. The
scope of these hearings was limited to the environmental effects of spent fuel reprocessing and radioactive
waste management. Environmental impacts of the entire fuel cycle are being studied by the NRC staff in a

program to revise an'd update fuel cycle impact values,
including radon releases from uranium mining andmilling and technetium releases from reprocessing andwaste management. This program has not yet reached
the stage where specific further amendments to Table
S-3 are ready to be proposed. Because radon andtechnetium releases are not now included in Table S-3,
the amount and significance of these releases may belitigated in individual reactor licensing proceedings.
In prom~lgating the new S-3 rule, the Commission
noted that the matters set out in Table S-3 do not, bythemselves, convey the environmental significance ofuranium ,fuel cycle activities. The focus of interest is
the health effects resulting from the effluents set out inTable S-3. Accordingly, the Commission directed thestaff to prepare for public comment in a further
rulemaking an explanatory narrative addressing thesignificance of fuel cycle releases. Pending adoption ofsuch a narrative, the Commission directed the NRC
staff to continue presenting in individual proceedings
an evaluation of dose commitments and health effects
from fuel cycle releases. In these proceedings, the staff
will also address economic, socioeconomic, and possible cumulative impacts of fuel cycle activities and suchother impacts as may reasonably appear to havesignificance for NEP A purposes.
The Commission stressed that the fuel cycle rule has
a limited purpose. It applies only to environmental
cost-benefit balances considered in reactor licensing
proceedings and is in no way intended to be a tool forchoosing among alternative uranium fuel cycle technologies.

Three Mile Island Unit 1
At the time of the March 28, 1979 accident at Unit 2of the Three Mile Island nuclear station, Unit 1 wasshut down for refueling. Unit 1 is essentially identicalto Unit 2, and is owned and operated by the same
licensee. During the period immediately after the accident, the licensee was instructed by the NRC staff not
to resume operation of Unit 1 pending staff approval.
On July 2, 1979, the Commission ordered that thefacility must remain in a cold shutdown condition
until further order of the Commission, and that ahearing must precede restart.
On August 9, 1979 the Commission specified in an
Order the issues to be considered in that hearing andthe procedures to be applied. The Order provided thatsatisfactory completion of certain "short-term actions"and reasonable progress toward completion of certain
other "long-term actions" are required to provide
reasonable assurance of adequate protection of the
public health and safety. The short-term actions included a number of mechanical and technical changes
that have been required of all similar nuclear
facilities. The Order also required other short-term actions resulting from: (1) potential interaction between
Unit 1 and the damaged Unit 2; (2) questions about

==================================================261
the management capability of the TMI licensee inview of the accident at Unit 2; and (3) recognized deficiencies in the licensee's operating procedures andemergency plans.
The Commission formed an Atomic Safety andLicensing Board to conduct a hearing mandated by
the August 9 Order and to determine whether the
short-term and long-term actions specified in theOrder are necessary and sufficient to ensure safe
operation of Unit 1. The Board was instructed to proceed expeditiously in conducting a fair and thoroughhearing and in arriving at a decision. Finally, theAugust 9 Order also provided a special procedure for
expedited Commission determination of whether aboard decision authorizing resumption of operation
should become effective pending completion of Commission review of that decision.

Offshore Power Systems
On September 14, 1979 the Commission issued aMemorandum and Order in the Offshore Power
Systems proceeding which affirmed a decision by an
Atomic Safety and Licensing Appeal Board that theprobability and consequences of a "Class 9" accident
are proper subjects for consideration in the Commission's environmental analysis of Offshore Power
System's application for a manufacturing license tobuild floating nuclear power plants. The
Commission's decision was based on the narrowground that it was not apparent that floating nuclear
power plants had been considered in the formulationof the accident classification system under which it
was determined that Class 9 accidents need not be considered. Further, the NRC staff had already preparedan environmental impact statement analyzing theClass 9 issue for the OPS proceeding and that statement called for imposition of specific licensing conditions to mitigate the consequences of such an accident.
Given those facts, it was felt that there would be a conflict with the philosphy of NEPA for the Commission
to blind itself to the existence of the stafrs analysis or
to refuse to consider it in licensing floating nuclear
power plants.
The Commission did not directly address thebroader question of consideration of Class 9 accidents
for land-based plants, but it instructed the NRC staff
to move ahead with completion of a long-dormant
rulemaking on that subject. The Commission also
directed the staff to prepare interim guidance in thisarea pending completion of that rulemaking, and tobring to the Commission's attention any individual
cases in which the staff believes the environmental
consequences of Class 9 accidents should be consideredas part of the licensing process.

Waste Confidence Rulemaking
On October 25, 1979, the Commission published a
notice of proposed rulemaking in the Federal Register

which announced the Commission's intent to conduct
a reassessment of its degree of confidence that highlevel radioactive wastes produced by civilian nuclear
facilities will be safely disposed of, to determine whensuch disposal will be available and whether such
wastes can be safely stored until they are safely disposed of. The rulemaking was initiated in response to thedecision in State oj Minnesota v. NRC, 602 F.2d 412(D.C. Cir. 1979), but it is also a continuation ofprevious proceedings conducted by the Commission in
the waste area.
The notice described the hybrid rulemaking procedures the Commission has determined to use, including the possibility that a legislative type hearing
may ultimately be held before the Commission itself.
The preliminary stages of the proceeding will be conducted by a presiding officer. To aid the participantsin the proceeding, the NRC staff will compile an ex- .
tensive data bank on the subject of high-level waste,
supplemented by an extensive bibliography. The tentative schedule announced by the Commission would
lead to promulgation of a final decision on the waste
confidence question in early 1981.

Bailly Short Pilings Petitions Denied
On December 21, 1979, the Commission denied two
petitions relating to the installation of pile foundations
for the Bailly generating station (Nuclear 1), locatedon the shores of Lake Michigan, near Gary, Indiana.
Commissioner Bradford dissented, and Commissioner
Gilinsky filed separate views.
The Bailly licensee, Northern Indiana Public Service Company (NIPSCO), received its construction
permit in 1974. In 1978, it submitted to the NRC staff
a plan detailing its intent to install foundation piles tobe embedded in the glacial lacustrine deposits
underlying the site. The petitioners asserted that thisplan represented a change from the plan approved at
the construction permit proceeding, and that a construction permit amendment (with attendant rights toa hearing) was therefore required.
Before reaching a decision on the petitions, theCommission solicited the views of the parties and of
the Advisory Committee on Reactor Safeguards. In
particular, the Commission asked the ACRS to assess
the significance, from the standpoint of engineeringand of safety, of a decision to use pilings embedded inglacial lacustrine deposits, rather than set on bedrockor in the glacial till just above bedrock. By letter of July 16, 1979, the ACRS replied that the use of shorter
pilings was not a significant design change from thestandpoint of engineering, would not require significant alteration of other aspects of the design, andwould not affect the safety of the facility, provided
that stated conditions were observed.
In denying the petitions, the Commission stated that
it found the short pilings plan to be less a change froman earlier design than a proposed resolution of a mat-262

ter that at the construction permit stage had properly
been left for later resolution. The Commission pointed
to references in the record indicating that resolution of
the pilings issue was contingent on pile tests to be conducted only after the issuance of the construction permit. The Commission stated that in reaching its decision as to whether to order a hearing, it had addressed
the safety issue, on which question the views of the
ACRS were particularly useful. The Commission concluded that as a matter of law, no construction permit
amendment was needed, and that a discretionary
hearing would serve no useful purpose. The Commission explicitly did not address the question of whattypes of design changes require construction permit
amendments.
Writing in dissent, Commissioner Bradford took theposition that the hearing record established a clear
commitment by the licensee to drive pilings to
bedrock; thus a construction permit amendment wasneeded as a matter of law. Commissioner Gilinsky, ina statement of separate views, concluded that in viewof the ACRS judgment as to the significance of thechange to shorter pilings, no construction permit
amendment was needed, but he would have favored
granting a hearing in any case, owing to the novelty of
the issue and the possibility that a hearing would
develop additional information.
(The Commission: s decisions on export licensing
cases are discussed in Chapter 9.)

JUDICIAL REVIEW
Significant Cases
Westinghouse Electric Corporation v. NRC (3rdCir., Nos. 78-1188, 78-1189).
Exxon Nuclear Company, Inc. v. NRC (9th Cir.,
No. 78~1403) (3rd Cir., No. 78-1840) (dismissed by Exxon, August 11, 1978).
Allied-General Nuclear Service v. NRC (D.C. Cir.,Nos. 78-1144,78-1422).
Scientists and Engineers for Secure Energy, MidAtlantic Legal Foundation, and Capital Legal Foundation v. NRC (3rd Cir., No. 78-1204).
This series of cases challenged the Commission's
December 23 order which terminated GESMO andrelated proceedings. The cases were consolidated in
the Third Circuit for argument and decision. Petitioners argued that completion of an EIS wasnecessary to terminate the proceedings, that the Commission showed too great a deference to the President's
foreign policy judgments, and that the Commission is
obliged to pass upon all license applications under
Atomic Energy Act standards. On April 19, 1979, theThird Circuit affirmed the Commission's decision terminating its GESMO review. In doing so, the court
agreed that the Commission could rightly give great
weight to the views of the President on foreign policy

issues, was empowered to freeze license proceedings
when considering overall policy issues, and had correctly determined that NEPA's EIS requirements did
not extend to a non-merits freeze decision. 598 F.2d
759.
Seacoast Anti-Pollution League, et al. v. NRC (lst
Cir., No. 78-1172).
This NEPA case was brought by two environmentalgroups challenging the consideration of alternative
sites in connection with the Seabrook application. On
May 30, 1979, the First Circuit affirmed the NRC
decision that the alternative site investigation for the
Seabrook facility complied with NEPA. 598 F. 2d
1221.
The only Seabrook case that was still pending at theend of the report period was the challenge to use of the
S-3 rule. (New England Coalition v. NRC, 1st Cir.,
No. 76-1525.) The First Circuit is apparently awaiting
the outcome of the S-3 litigation in the D.C. Circuit.
People of the State of Illinois v. NRC, et al. (7thCir., No. 78-1171).
Illinois petitioned the Court of Appeals to review
the denial of its request for enforcement action (underCommission regulation 10 CFR 2.206) on the General
Electric facility at Morris, Ill.
Petitioner alleged that the Morris facility had been
"converted" to long-term storage for radioactive waste
without preparation of an impact statement andwithout an evidentiary hearing. On January 10, 1979,
the Seventh Circuit upheld NRC's denial of a petition
for relief under 10 CFR 2.206, holding that NRC was
not required to hold a formal hearing before it could
rule on an enforcement petition. 591 F.2d 12.
Mississippi Power and Light Company, et al. v.
NRC, et al. (5th Cir., No. 78-1565).
Nuclear Engineering Company v. NRC, et al. (5thCir., No. 78-1871).
Chem-Nuclear Systems v. NRC, et al. (5th Cir., No.
78-2200).

A number of utilities sued the NRC on its February
9, 1978 license fee rule. The utilities alleged that NRC
exceeded its statutory authority in setting the fees.
They sought a declaration that the fee schedules are invalid, a suspension of collections in the interim, and a
refund of all fees collected under the rule and its 1973predecessor. The Fifth Circuit affirmed the NRC
schedule generally and as against each' specificchallenge on August 24, 1979. 601 F.2d 223. The
utilities have petitioned the Supreme Court to hear
this case. (See also Chapter 14.)
State of Minnesota, By the Minnesota Pollution
Control Agency v. NRC and the United States (D.C.
Cir., No. 78-1269).
New England Coalition on Nuclear Pollution v.
NRC (2d Cir., No. 78-4103).

==~============================================263

Minnesota sought review of the Appeal Board's decisio~ in ALAB-455 (Norther States Power Company),

whICh authorized expanded spent fuel storage at theapplicant's Prairie Island facility. The New England
Coalition also sued to review ALAB-455, claiming
spent fuel storage at the applicant's Prairie Island
facility. The New England Coalition also sued toreview ALAB-455, claiming in connection with aspent fuel pool expansion proceeding. On May 23,
1979, the D.C. Circuit remanded to the Commission
the Appeal Board's decision in ALAB-455 for the Commission to decide whether, in light of recent events
such as the report of the Interagency Review Group on
Nuclear Waste Management, there is reasonableassurance that off-site storage would be available forspent fuel at the end of the license period. 602 F.2d
412. The proceedings ordered by the D.C. Circuit
have been initiated by the Commission.
Ft. Pierce Utilities Authority of the City of Ft.
Pierce, et al. v. United States, et al. (D.C. Cir., Nos.
77-2101, 77-1925).
In Ft. Pierce Utilities, petitioners asked the Court ofAppeals to review two related Commission actions denying an antitrust hearing. Petitioners argued that a<:om~ission antitrust ~evie~ may be initiated at any
tIme, mdependent of hcensmg reviews. On March 23,
1979, the Court of Appeals affirmed the NRC determination that the Section 105 antitrust amendments of
19~0. ~o the Atomic Energy Act exempted existingfacIhtIes then under construction from antitrust
review at the operating license stage. 606 F.2d 986.
The court did not reach the broader issue of whether
those amendments were the sum total of the Commission's antitrust authority. The Supreme Court declined
to review the decision on October 1, 1979. 100 S. Ct.
83.
Westinghouse Electric Corp. v. Hendrie (D.D.C.,
No. 79-2060, on appeal, D.C. Cir., No. 79-2069).
Westinghouse Electric Corp. v. Vance (D.D.C.,
No. 2110, on appeal, D.C. Cir., No. 79-2070).
The Westinghouse Corporation sued the NRC andthe Department of State, alleging unreasonable delayin the processing of its licenses to export a reactor andcomponents to the Philippines. On August 31, Judge
June Green denied the Westinghouse motion for injunction, while bypassing NRC's jurisdiction
arguments, and found that the NRC delay was notunreasonable given the important ·health and safety
considerations implied by the application. After thedenial of its motion, Westinghouse appealed to the
D. C. Circuit and is in the process of having the record
certified. (See Chapter 9.)
Natural Resources Defense Council, Inc., et al. v.
Nuclear Regulatory Commission, et al. (U.S.C.A.,
D.C. Cir., No. 74-1385, and related cases Nos.
73-2266, 73-1776, 73-1867, 74-1586, and 77-1905).

These cases are challenges to the Commission's fuelcycle rule (Table S-3) on remand from the Supreme
Court. They have been held in abeyance pending completion of the Commission's rulemaking.
State of New York v. NRC (D.C. Cir., No.
79-2110).
"Natural Resources Defense Council, et al. v. NRC(D.C. Cir., No. 79-2131).
These two cases are challenges to the new Commission fuel cycle rule (Table S-3) which became effective
on September 4, 1979.
City of Lancaster v. NRC (D.D.C., No. 79-1368).
Susquehanna Valley Alliance v. NRC (M.D. Pa.,No. 79-658).
In separate actions filed in May 1979, plaintiffs sued
to bar use of the EPICOR-II demineralizer facility anddischarge of waste water from the Three Mile Island
nuclear power facility in Pennsylvania pending completion of an environmental impact statement andlicense amendment proceedings.
On October 12, 1979, the District Court in Harrisburg dismissed the Susquehanna Valley Alliancecomplaint for lack of subject matter jurisdiction,noting that plaintiffs had made no effort to employ theprocedures set out in 10 CFR 2.206, and concluded
that they had therefore failed to exhaust their administrative remedies. The Court also found thatneither of the two exceptions to the exhaustion doctrine were met in this case since plaintiffs had notshown that recourse to the Section 2.206 remedy
would be futile, nor had they shown that NRC had
violated a clear, nondiscretionary legal duty. Finally
the District Court noted that, should plaintiffs seek
agency relief pursuant to Section 2.206, any final NRC
decision on their request would be reviewable exclusively in the courts of appeals.
An appeal to the Third Circuit Court of Appeals
followed. On October 18, the Third Circuit denied
SVA's motion for injunction pending appeal. The case
was argued on the merits in November.
In City of Lancaster, the NRC's motion to dismisshas been set for argument in January 1980.
Three Mile Island Litigation (M.D. Pa. No.
79-432).
'
These are consolidated cases seeking damages for injuries claimed to arise from the TMI Unit 2 accident.
The Commission, which is not named as a defendanti~ participating as a friend of the court in the litiga~
hon. The cases are at a preliminary discovery stage
and have not yet been certified as a class action.
Friends of the Earth v. NRC (9th Cir" No.
79-7311).
Petitioner seeks review of the NRC decision torestart the Rancho Seco facility after it was shut down

264

for modifications following the accident at Three Mile
Island. Briefing is now underway.
.
Akron, Canton & Youngstown Railroad Company,
et al. v. Interstate Commerce Commission, et al. (6th
Cir., No. 78-3425).
Twenty-two railroads petitioned the Sixth Circuit to
set aside an order of the Interstate Commerce Commission (ICC) in five consolidated cases. The railroads
seek a declaration that they are not common carriers ofhighly radioactive nuclear materials. The NRC filed alimited appearance before the ICC to argue that ICC
lacked jurisdiction to examine the health and safety
aspects of nuclear materials transportation. The NRC
moved to intervene in the case. The case was argued in
February 1979 and is awaiting the Court's decision.

Natural Resources Defense Council, Inc., et al. v.
Robert C. Seamans, Jr., et al. (D.D.C., No. 76-1691,
on appeal, D.C. Cir., Nos. 78-1576, 78-1698).
Natural Resources Defense Council, Inc. v. NRC(D.C. Cir., No. 77-1489).
In re Robert W. Fri, Acting Administrator of ERDA
(D.C. Cir., No. 77-121D).
NRDC and other environmental groups sued ERDA
and NRC seeking to block construction of the waste
tanks intended for the Hanford and Savannah Riverfacilities. The complaint alleged that ERDA (now the
Department of Energy) had failed to comply withNEPA by not issuing an environmental impact statement for the waste tank construction, and had failed
to obtain licenses from NRC under Section 202(4) of
the Energy Reorganization Act-. NRC was named adefendant because plaintiffs sought a declaratory
judgment that NRC has licensing authority in thismatter.
On May 8, 1978, the District Court Judge issued a
34-page opinion upholding NRC's position that it lacks
licensing authority over the Hanford and SavannahRiver storage tanks, but found that DOE erred in notpreparing project-specific environmental impact
statements for the waste tanks. Cross-appeals were filed. On August 17, 1979, the Court of Appeals reversed
the District Court's determination that it had subject
matter jurisdiction to review the NRC decision, finding that the NRC decision was a licensing determination and thus exclusively reviewable in the courts ofappeals. The D. C. Circuit affirmed the Commission's
determination that, pursuant to Section 202(4) of theEnergy Reorganization Act, it lacked licensing
jurisdiction over 22 ERDA high-level radioactive
waste tanks under construction at Hanford and Savannah River. 606 F. 2d 1261.

Porter County Chapter of the Izaak Walton Leagueof America, et al. v. NRC (D.C. Cir., No. 78-1556).
People of the State of Illinois v. NRC (D.C. Cir.,
No. 78-1599).

The City of Gary, Indiana v. NRC (D.C. Cir., No.
78-1560).
The Lake Michigan Federation v. NRC (D.C. Cir.,
No. 78-1561).
These petitions sought review of the Commission'sApril 20, 1978 decision affirming the denial by the
Director, NRC Office of Nuclear Reactor Regulation,of a 2.206 enforcement request relating to the Bailly
Generating Station. The cases were consolidated onJune 23. The D.C. Circuit, on September 6, 1979, affirmed the Commission's denial of the petitioners' request for remedial relief under 10 CFR 2.206. In its
decision, the Court gave strong support to a wideranging role for the NRC staff in nuclear regulation, to
the Commission's procedures for responding to 2.206
requests, and to the operating license review as theusual forum for addressing issues that arise during
plant construction that relate to whether the completed plant can be operated safely. 606 F.2d 1363.

Hunt, et al. v. NRC, et al. (N.D. Okla., No.
79-C-122-C, ajj'd., 10th Cir., No. 79-1647).
Plaintiffs, observers at the Black Fox proceeding,challenged under the Sunshine Act a licensing board's
ability to conduct a closed session to discuss a reportcontaining proprietary data. The District Court
denied a temporary restraining order, and NRC filed amotion to dismiss the action on February 26, 1979.
The District Court agreed with NRC's interpretationof the Sunshine Act-that it applied to quorums of theCommission and not sessions of the NRC licensing
board. 468 F.Supp. 817. On August 24, 1979, plaintiff
appealed to the Tenth Circuit. The Court of Appeals
for the Tenth Circuit affirmed the lower court's decision, holding that "sessions of the ASLB are notcovered by the Sunshine Act . . . . " No. 79-1647,
November 23, 1979.

Concluded Cases
Basdekas v. NRC, et al. (D.D.C., No. 78-465.)
On March 17, 1978, an NRC employee sued to compel disclosure of documents under the FOIA andPrivacy Act. The documents were an investigative
report of the Commission's Office of Inspector and
Auditor and two memoranda from the Office of theGeneral Counsel to the Commission, relating to a 1976
OIA investigation. The NRC asserted that portions of
the documents are exempt from disclosure under Exemptions 5 and 6. On December 26, 1978, Judge
Green ordered the release of an OGC memorandum
but upheld the NRC's claims of exemption for theother two, including the statements of witnesses provided under a pledge of confidentiality. NRC settled
the remaining issues in the case.

======~===========================================265

Martin Hodder, et al. v. NRC, et al. (D.C. Cir.,Nos. 76-1709, 78-1149) (S. Ct., No. 78-1652).
Petitioners brought two petitions for review of theadministrative decisions in this case, challenging the
NRC construction permit for St. Lucie Unit 2 on theeast coast of Florida. The three issues raised by thecase were whether the Commission's treatment ofClass 9 accidents satisfied NEPA, whether the St.
L~cie Unit 2 site complied with Part 100 of the Commission's regulations, and whether the comparison ofalternative sites was sufficient to support a decision tobuild the reactor at St. Lucie. In a memorandum decision on December 22, 1978, the D.C. Circuit upheld
the NRC's decision to permit construction of the St.
Lucie 2 reactor, holding that NRC was not legally required to evaluate Class 9 accidents for NEPA purposes, that the Part 100 siting regulations were properly applied, and that the investigation of five alternative sites satisfied NEPA. 589 F.2d 1115 (Table). On

Two thirds of the Commission meetings in fiscal year 1979 wereopen to the public, and many drew large audiences, particularly
after the Three Mile Island accident. Shown is part of an attentive

October 1, the Supreme Court denied the petition for
writ of certiorari. 100 S. Ct. 55. The Court subsequently denied a petition for rehearing. 48 U.S.L.W.
3357, November 27, 1979.
Chauncey Kepford v. NRC (D. C. Cir., No.
78-1933).
On September 21, 1978, petitioner sued the NRC forreview of ALAB-480, an appeal board decision which
established a procedure to conduct evidentiary hearings on the radon issue in cases pending before theboard. Petitioner sought review only insofar as
ALAB-480 affects the Three Mile Island proceeding.
On May 14, 1979, the court dismissed this petition toreview ALAB-480 as not a final order for the purposeof judicial review.
Pennsylvania Power & Light Co. v. Gulf Oil Co.
(Ct. of Common Pleas, Lehigh County, Pa., Equity
Action No. 75-453).

crowd at NRC headquarters during a briefing of the Commission
by one TMI investigating group.

266

On October 16, 1978, Gulf Oil Company subpoenaed NRC records in its effort to defend a breachof contract effort over uranium for the Susquehanna
station brought by the applicant, Pennsylvania Power& Light Co. NRC treated this request as if filed under
the FOIA, and made some documents publicly available. At NRC's request, Gulf voluntarily withdrew its
subpoena and its discovery requests since all
documents in their possession are now in the PublicDocument Room and available to them.
Robert Fendler v. NRC;, et al. (D. Ariz., No.
79-239-PHX) .
In a near duplicate of a case brought several years
earlier, plaintiff sued the NRC to compel preparationof a new environmental impact statement for the Palo
Verde plant. This case was dismissed on August 6,
1979.
Virginia Sunshine Alliance v. NRC, et al. (D.D.C.,
No. 79-1989, on appeal, D.C. Cir., No. 79-2060).
On July 31, 1979, plaintiff sued to block the shipment of spent fuel from foreign research reactors
through Portsmouth, Virginia, alleging the possibilityof sabotage to the shipment. On August 3, federal
District Court Judge Penn denied plaintifFs request to
preliminarily enjoin spent fuel shipments through Norfolk, Virginia, finding that the Commission's new
safeguards rule provided adequate protection against
sabotage threats. The denial of the injunction is now
on appeal.
Township of Lower Alloways Creek v. NRC(D.N.J., No. 79-1129.)
County of Ocean v. NRC (D.N.J., No. 79-1800).
Plaintiffs sought to enjoin expansion of the spent
fuel storage pools at the Salem, Oyster Creek, andForked River facilities. The cases were consolidated
for the purposes of argument, which was held in
Camden in September 1979. On December 11, 1979,
NRC's motion to dismiss these consolidated cases forlack of subject matter jurisdiction, failure to exhaustadministrative remedies, and (in the case of Forked
River) the absence of a valid case or controversy, wasgranted.
United States of America and the Trustees ofColumbia University in the City of New York v. Cityof New York, et al. (S.D.N.Y., 77 Civ. 3485, on appeal, 2d Cir., No. 79-6023).
The United States, on behalf of NRC and ERDA
(now DOE), and Columbia University, filed a joint
complaint against the City of New York asserting thatthe city's refusal, on radiological health and safety
grounds, to permit an NRC-licensed reactor to operateviolates the supremacy clause of the United States
Constitution. The complaint sought a declaration andinjunction against enforcement of section 105.107(c)of the City's Health Code which purports to require a

City radiological health and safety review and permit
for operation of an NRC-licensed reactor. NRC contended that the Atomic Energy Act preempts local
authorities from regulating the health and safety
aspects of nuclear reactor operation.
On December 27, 1978 the District Court invalidated New York City's requirement of a separate
City permit for the Columbia Triga Reactor, findingthat New York City's proposed reassessment ofradiological safety conflicts with the Atomic Energy
Act's exclusive grant of authority in this field to theFederal Government. 463 F. Supp. 604. An appeal
was filed with the Second Circuit. After Columbia
cancelled its plans for the Triga Reactor, the Second
Circuit ordered the District Court to dismiss the caseas moot.
Mid-America Coalition for Energy Alternatives,
Inc. v. NRC (D.C. Cir., No. 78-1294).
Petitioner sought review of the appeal board's
March 9, 1978 decision in ALAB-452 which affirmed
the licensing board's authorization of a construction
permit for Wolf Creek Generating Station, Unit 1. On
April 24, petitioner sought a stay from the Court ofAppeals, which NRC opposed. On July 6, the Court ofAppeals denied the stay but ordered that the caseshould be set for argument as soon after the filing of
the NRC brief as business permits. On January 15,
1979, the D.C. Circuit affirmed NRC's decision to permit construction of the Wolf Creek Station. The Court
held that petitioner's proposed alternative was raised
too late and that the NRC gave proper independentconsideration to the alternatives in its environmental
statement. 590 F.2d 356 (Table).
Jeannine Honicker v. Joseph Hendrie, Chairman,
NRC, et al. (M.D. Tenn., Civ. No. 78-3371-NA-CV)(6th Cir., No. 78-1405, on appeal, 6th Cir., No.
79-1132) (S. Ct., No. 79-710).
Plaintiff sued the NRC for injunctive relief, alleging
that the NRC had permitted nuclear power reactors tooperate while underestimating the magnitude ofhealth effects of the nuclear fuel cycle. Plaintiff sought
revocation of all licenses and dismantling of all existing fuel cycle facilities. The District Court, on
September 6, denied a temporary restraining order.
Plaintiff appealed the denial of the temporary
restraining order to the Sixth Circuit. Not receiving an
immediate decision, she brought her request to theSupreme Court. Mr. Justice Stewart denied her request on September 14, 1978. Subsequently, the Sixth
Circuit denied emergency relief on November 6.
Honicker v. Hendrie, et al., 6th Cir., No. 78-1405.
On NRC's motion, the District Court dismissed thecase for lack of jurisdiction on January 16, 1979. 465
F. Supp. 414. The District Court found that plaintiff
had failed to exhaust her available administrative
remedies and that, in view of the scientific controversy, the case was appropriate for invocation of the doc-

267

trine of primary jurisdiction. Then, because review ofsuch an NRC decision was vested exclusively in the
courts of appeals, the District Court dismissed thecase. The plaintiff appealed to the Sixth Circuit
which, on August 7, 1979, affirmed the District
Court's order of dismissal, "[f]or the reasons set forthin the district court's thorough and well-reasoned
memorandum opinion:' 605 F.2d 556. A petition for
writ of certiorari is pending.
Gilbert, et al. v. NRC (S.D. Texas, No. H-78-2192).
Plaintiffs challenged various aspects of the NRC
procedures to be followed in proceeding on the Allens
Creek application for a construction permit. The NRC
moved to dismiss this case on grounds of lack of subject
matter jurisdiction. On January 23, 1979, the District
Court entered a brief order dismissing plaintiffs' complaint, finding, in effect, that petitioners had not exhausted their administrative remedies and that nofinal order had been entered in the case.
Tibor Fischer v. Nuclear Engineers, et al. (E.D.
N.Y., Civ. No. 78C-259).
On March 7, 1978, plaintiff sought "discharge andrelease from scientific talks and burning pressures."
On NRC's advice, the U.S. Attorney moved for
dismissal on the basis that the plaintiff failed to state acause of action. On May 14, 1979, the District Court
granted the motion to dismiss this complaint.
Radiation Technology v. NRC (D.N.J., No.
79-753).
Plaintiff sought money damages under the Federal
Tort Claims Act for costs flowing from the suspensionof this materials license. NRC's response alleged thatcounts 1 and 2 of the complaint were time-barred
under the Tort Claims Act, and disputed the facts of
the remaining claim. Judge Stern granted NRC's motion to dismiss counts 1 and 2 on statute of limitations
grounds; the remaining claim is in the process of being
settled.
Natural Resources Defense Council, Inc., et al. v.
NRC, et al. (D. New Mexico, No. 77-240-B).

This case, brought by the Natural Resources
Defense Council, challenged operations of a uranium
milling facility in New Mexico. On May 3, 1977,
NRDC, the Central Clearinghouse of New Mexico,and two individuals filed suit against NRC and the
New Mexico Environmental Improvement Agency(NMEIA), to enjoin operations of United Nuclear's
Church Rock Mill, which NMEIA licensed, alleging
violations of NEPA and the Atomic Energy Act. The
gist of the complaint was that neither NRC nor New
Mexico prepared an environmental impact statement
for the Church Rock Mill. Plaintiffs cQntended that
New Mexico, as signatory to a Section 274 State Agreement to regulate redioactive materials, exercisedFederal power and therefore must comply with NEPA,

and the NRC's continuing review powers over State
programs constituted sufficient Federal involvement
to call for preparation of an environmental impact
statement (EIS). Second, plaintiffs argued that, inorder to comply with section 274, State programs mustbe "compatible" with the NRC program and that compatibility required preparation of an EIS where NRC
would prepare one in a non-Agreement State. NRC
currently prepares an EIS for each new milling licenseand first renewal. A similar petition was filed June 30,
1977, in the D.C. Circuit naming only NRC as a
respondent (No. 77-1570).
The D.C. Circuit on January 6, 1978, issued anorder which rejected NRDC's theory that New Mexico
as an Agreement State is exercising delegated Federal
power. The Court also found that NRDC's allegations
that the NRC has been "intimately involved" with thelicensing of Church Rock demonstrate, if true, onlyState-Federal cooperation rather than a final decisionmaking authority retained by NRC. The Court took noview on whether the New Mexico regulatory program
is compatible with the Federal regulatory framework.
That order brought the proceedings before the D.C.
Circuit to a close. 8 ELR 20153.
Motions for summary judgment were filed before
the District Court for the District of New Mexico bothby NRDC and by the NRC; trial was scheduled earlyin 1979. The Kerr-McGee Nuclear Corporation intervened before the District Court, after the June 15,
1978 decision of the Tenth Circuit, on the appeal of
Kerr-McGee and the American Mining Congress,
which reversed the District Judge's decision denying
them intervention. 10th Cir., 578 F.2d 1341.
The District Judge approved a stipulation on July
13, 1979, dismissing this case in light of the Uranium
Mill Tailings Radiation Control Act of 1978, whichchanged the regulatory control over the tailings.

Opened Cases
Thot-Thompson v. McVeagh (D. Md., No.
B-79-1703) .
Plaintiff sued defendant for damages alleged to be
the result of slander. Both are NRC employees. Defendant is plaintiffs supervisor and is alleged to haveslandered plaintiff by commenting upon plaintiffs job
performance. The NRC position is that the defendant
was acting within the scope of his employment when
he made statements about the plaintiff. A petition forremoval from State court to Federal District Court
was filed on September 13. The U.S. Attorney will
represent the defendant.
Gentry, et al. v. United States, et al. (N.D. Ala.,
No. CA79-L-5181-NE).
Plaintiff seeks money damages for injuries allegedly
caused by radiation exposure during the years
1954-1964 while plaintiff was working for Thiokol
Corporation on various government projects. The

268

Justice Department will have lead responsibility forthe defense of this action, and has filed a motion todismiss, or in the alternative, for summary judgment.
Kertis v. United States (W.D. Pa., No. 77-1259).
Plaintiff seeks recovery as administratrix of her husband's estate. Plaintiffs decedent, who died ofleukemia in 1974, was a worker in the Westinghouse
Cheswick facility engaged in repair of Navy submarine pumps. A similar lawsuit was dismissed in
1976 as plaintiff was limited to workmen's compensation against Westinghouse under State law. NRC
assists the U.S. Attorney.in responding to requests fordiscovery in this case. The NRC involvement is ex,tremely limited, since the specific NRC license includes only a limited amount of radioactive material.
Lawrence Skinner v. NRC (N.D. Calif., No.
C-79-1231-WAI).
Plaintiff seeks tort relief from the NRC for injuries
alleged to have resulted from atomic weapons testingin the 1950s. This case is being handled through the
U.S. Attorney's office. NRC's position is that DOE,
and not NRC, is the responsible agency here. Even if
NRC were the proper defendant, plaintiff has not exhausted his administrative remedies. A motion todismiss was filed on August 9, 1979.
Broudy v. United States (C.D. Calif., No. 79-02626
LEW(GX)).
This is a tort action for damages against the UnitedStates and various government agencies for wrongfuldeath of Broudy, a Marine participant in an AEC
weapons test in Nevada in 1957. The NRC position issimilar to that taken in Punnett and Skinner (above).
The NRC is not responsible in any way for weapons
programs, that function being transferred to the
Department of Energy. The Justice Department ishandling this defense.
Life of the Land v. Adams (D. Hawaii, No.
79-0249).
Plaintiffs challenged the transport of two shipmentsof spent fuel from Japan through Hawaiian waters andthe port of Honolulu, seeking preparation of an environmental impact statement and compliance withNEPA and the Ports and Waterways Safety Act. The
application for injunction on the first of these
shipments was denied on June 7, 1979, and upheld by
the Ninth Circuit on June 8, 1979. The governor closed
the port to both shipments. One was permitted torefuel at Pearl Harbor on an emergency basis; theother refueled in non-Hawaiian waters. Because no
more shipments were scheduled, the Justice Department filed a motion to dismiss on grounds the case was
moot.
Won-Door Corporation v. United States (Court ofClaims, No. 109-79L).
Won-Door sued the United States for compensation

for an alleged taking of its property by virtue of radon
contamination from the adjoining Vitro urarIium milltailing site. The Department of Justice is handling thedefense of this action. The NRC advised the JusticeDepartment of its opinion that no taking of the plaintiffs property occurred through regulatory actions.
The answer to the petition was filed June 11, 1979.
NRC responded to interrogatories in August 1979.
State of Illinois v. General Electric (N.D. 111., No.
79-C-1427) .
The State seeks a declaration that General Electric is
the sole and exclusive owner of the Morris site and thatthe prior arrangements for perpetualcare of the site by
the State under Illinois law are invalid. The NRC is anamed defendant in this lawsuit although the principal dispute concerns General Electric and the State.
The Illinois Attorney General seeks to invalidate theIllinois Radioactive Waste Act as preempted by theFederal Atomic Energy Act. It is NRC's position thatthe acts are consistent, and the agency filed a motion
to dismiss in July 1979. As to claims that the law is invalid under the Illinois constitution, the NRC would
defer to the Illinois State courts now considering thisissue. The motion to dismiss is pending.
Upper Skagit Indian Tribe, Suak-Suiattle Indian
Tribe and Swinomish Tribal Community v. NRC andUnited States of America (D.C. Cir., No. 79-2277).
On October 26, 1979, three American Indian tribes,
the Upper Skagit Indian Tribe, the Suak-Suiattle Indian Tribe, and the Swinomish Tribal Community
sued the NRC concerning Appeal Board decisions denying their petition to intervene, which was filed 3-112
years late into the Skagit construction permit proceeding. NRC has moved to hold the petition inabeyance pending the outcome of the administrative
proceedings.
Susquehanna Valley Alliance v. NRC (3rd Cir., No.
79-2800).
On December 13, 1979, petitioner filed a request forreview of the October 16, 1979, order directing use of
the EPICOR-II decontamination system at TMI-2.

Pending Cases
Minnesota Environmental Control Citizen's
Association, et al. v. Atomic Energy Commission, etal. (D. Minn., No. 4-72-109).
Plaintiff, a citizen's association, sought to enjoin further development and operation of Northern States
Power Company's Monticello and Prairie Island
facilities on the ground that the Prairie Island construction permit and the Monticello provisional
operating license were issued without preparation ofan an environmental impact statement. On July 28,
1972, the District Court issued an opinion refusing toenjoin the construction or provisional operation, but

==================================================269
holding that before full operating permits for thesefacilities could be granted, a full NEPA review was required. The Court retained jurisdiction over the matter to ensure that such a review was performed. During the past six years, the Commission has undertaken
this environmental review, and both licensing proceedings are nearing completion. Once these are completed, NRC intends to move to dismiss the complainton the grounds of mootness, as well as the statutory
mandate that only a court of appeals shall review finalorders of the Commission.
West Michigan Environmental Action Council, Inc.
v. AEC, et al. (W.D. Mich., No. G-58-53).
Citizen-group plaintiffs sought an injunction
against increased use of mixed oxide fuel in Consumers
Power Company's Big Rock Point power reactor. In
June 1974, the Court placed the case in abeyance pending the outcome of the GESMO proceedings and
NRC review of Executive Branch comments. The utility has not pressed its application nor prepared the required environmental report, so the case may eventually be moot.
Chauncey Kepford v. NRC, et al. (D.C. Cir., Nos.
78-1160,78-2170).
In No. 78-1160, petitioner sued the NRC to stay
operation of the Three Mile Island, Unit 2 facility,
primarily because of the level of radon-222 releases
from tailings produced in uranium mining and milling. On March 8, 1978, the Court denied petitioner's
motion for a stay. On March 22, 1978, the Court, on
its own motion, held further review in abeyance pending completion of the administrative appeals andordered NRC to file periodic status reports with the
Court.
In No. 78-2170, petitioner sued the NRC on
November 13, 1978, for review of the Commission's
affirmation of an appeal board decision that involved
all but two of the issues associated with the Three Mile
Island facility and which permitted its continued
operation. On November 30, 1978, the NRC moved tohold the petition for review in abeyance pending theoutcome of administrative hearings into one of theissues raised by the petitioner; that is, the probability
that a very large aircraft will crash into the reactor.
On May 11, 1979, the court granted the motion, andthis case is now being held in abeyance.
Ecology Action of Oswego, New York v. NRC, et al.
(D.C. Cir., No. 78-1855).
Petitioner sued the NRC to set aside the construction
permit for the Sterling nuclear facility. Petitioner
appeals from the denial of the application for a stay
before the appeal board. Briefing was completed on
March 14, 1979. Petitioner waived argument and thecase awaits decision on the briefs and record.
Detroit Edison Company, et al. v. NRC (6th Cir.,
No. 78-3187).

National Association of Regulatory Utility Commi<;sioners v. NRC (6th Cir., No. 78-3196).
These cases involved challenges to the Commission's
denial of a rule-making petition filed by the Detroit
Edison Company. Detroit Edison had requested that
the NRC amend its regulations to provide that NRC
lacked authority to require rerouting of transmission
lines associated with nuclear plants. This case involves
the same legal issue that the First Circuit decided
favorably to the Commission on June 21, 1978, inPublic Service Company of New Hampshire v. NRC,
No. 77-1419. 582 F.2d 77, cert. denied 99 S. Ct. 721.
The cases have been briefed and still await argument.
State of New York v. NRC (2d Cir., No. 75-4278).
Natural Resources Defense Council, Inc., et al. v.
NRC, et al. (2d Cir., No. 75-4276).
Allied-General Nuclear Service, et al. v. NRDC, etal. (Sup. Ct., No. 76-653).
Commonwealth Edison Company, et al. v. NRDC,.
et al. (Sup. Ct., No. 76-762).
Baltimore Gas & Electric Company, et al. v.
NRDC, et al. (Sup. Ct., No. 76-774).
Westinghouse Electric Corporation v. NRDC, et al.
(Sup. Ct., No. 76-769).
In these cases, New York State and certain citizen
groups sought review of the Commission's November
14, 1975 Federal Register notice, which set forth procedures for hearings on the Generic Enyironmental
Statement on Mixed-Oxide Fuel (GESMO) and outlined agency standards for licensing activities related tothe use of mixed oxide fuel prior to a Commission decision on wide-scale use of plutonium recycle. On May
26, 1976, the Court of Appeals for the Second Circuit
issued its decision upholding, in full, both the GESMO
hearing procedures and associated individual licensing
procedures. However, interim licensing (except for"experimental and feasibility purposes") was forbidden. The Supreme Court granted petitions for certiorari by a number of utilities and a manufacturer.
On December 23, 1977, however, the Commission
voted to terminate the GESMO proceedings.
In January 1978, the Solicitor General filed a suggestion of mootness on behalf of NRC with theSupreme Court, submitting that the Commission's
decision in December of 1977 on mixed oxide fuelrendered the Second Circuit's decision moot, and thatthe opinion should be vacated and the case remandedwith instructions to dismiss. On January 16, 1978, theSupreme Court vacated the Second Circuit's judgment
and remanded the case to the Second Circuit "to consider the question of mootness." Those cases are still
pending in the Second Circuit.
fohn Abbotts, et al. v. NRC (D.D.C., No. 77-624).
John Abbotts, the Public Interest Research Group,and the Natural Resources Defense Council, Inc.,

270

brought a Freedom of Information Act suit challenging an NRC decision to withhold certain safeguards
documents. The documents involved fall into three
categories: (1) records related to the NRC program foronsite review of SSNM facilities initiated in early 1976;(2) records concerning the NRC investigation andreview of conditions at the Nuclear Fuel Services
facility in Erwin, Tennessee, in late 1975 and early
1976; and (3) studies done for or related to NRC's
Special Safeguards Study and the Draft Safeguards
Supplement. Parties cross-moved for summary judgment in early 1978. After further review of the
documents by NRC, the parties narrowed the disputeto two small portions of two documents, concerning"baseline threat level" information, specifically contesting the proper classification of that information.
Supplemental cross-motions for summary judgment
have been filed in 1979. The Court must now decide
whether to review the documents in camera andwhether there is a valid exemption claim by NRC.
Natural Resources Defense Council, Inc. v. NRC, etal. (D.C. Cir., No. 77-1448).
On May 13, 1977, NRDC filed a petition for reviewof the NRC's March 14 Federal Register notice promulgating an interim rule quantifying the environmental effects of the uranium fuel cycle. On July 5,
1977, NRDC requested that the D.C. Circuit hold thecase in abeyance until the Supreme Court reaches adecision in the Vermont Yankee fuel cycle case. NRC
consented to that motion. On June 7, 1978, the D.C.
Circuit requested the parties' views on how to disposeof this other fuel cycle case. On June 27, 1978, NRC
advised the Court that the interim rule case should bedismissed or held in abeyance pending a challenge tothe final fuel cycle rule. The Court is holding the casein abeyance and considering a new scheduling order inlight of the issuance of final rule.
United States v. New York City (S.D.N.Y., No. 76
Civ. 273).
On January 15, 1976, the plaintiffs-the NRC,
ERDA, and Department of Transportation(DOT)-sought a judgment declaring a New York
City Health Code provision dealing with the transportation of nuclear materials through the city to be inconsistent with the Federal statutory scheme governing the transportation of hazardous materials. The
government's request for a preliminary injunction
against enforcement of the Health Code provision was
denied on January 30, 1976, the Court finding that no
irreparable injury would occur pending a decision on
the merits of the case. DOT has published regulations
under the Hazardous Materials Transportation Act(which became effective January 1977) which allow
interested persons to seek a ruling that a local ordinance is inconsistent with DOT regulations. On
February 28, 1977, Brookhaven filed its request forsuch a regulation with DOT, arguing that the city's

restrictions on shipping new and spent fuel were inconsistent with DOT's regulations. NRC and ERDA
(now DOE) wrote DOT in support of Brookhaven's
position. un April 4, 1978, DOT ruled that the city ordinance was not inconsistent with DOT policy, butthat a rule-making would be held to consider whether
regulations regarding the routing of nuclear materials
by road are warranted. DOE has asked the Depart~
ment of Justice to again bring a lawsuit. On FebruaryS, 1979, NRC advised the Department of Justice that it
did not favor such a suit.
Hope Punnett v. Jimmy Carter, et al. (E.D. Pa.,No. 79-29).
Plaintiff seeks notification to soldiers involved innuclear weapons testing programs and to their familiesof potential risks arising out of that program: The
Department of Justice is the lead agency in this action,and has pending a motion to dismiss on the basis ofsovereign immunity and no duty to notify. Discovery is
now taking place. Plaintiff amended its complaint toinclude a request for damages; but the class action
nature of the complaint has not yet been certified.
Plaintiffs motion for a preliminary injunction was
denied on March 30, 1979.
Peshlakai v. Schlesinger, et al. (D.D.C., No.
78-2416) .
Plaintiff alleges that there is both a national and aregional program to develop and process uranium"yellow cake" being carried out in violation of NEPA.
The Department of Justice is coordinating the defenseof this lawsuit. On September 5, 1979, the Court
denied plaintiffs motion for a preliminary injunction
on the Mobil in situ pilot project.
Charles Eason v. NRC (D.D.C., No. 79-0845).
Plaintiff in this FOIA lawsuit seeks access to theMedia Monitor, a copyrighted document produced byan outside contractor for the NRC. This case has progressed through discovery and is pending before the
Court on NRC's motion for summary judgment.
NRC's position is that plaintiff was never denied accessto the document since it is available in the NRC PublicDocument Room for inspection. Such availability does
not constitute a denial under the FOIA.
Commonwealth of Kentucky v. NRC (D.C. Cir.,
No. 78-1369).
The Commonwealth of Kentucky seeks review of
the appeal board's decision of February 16, 1978, andof a licensing board decision of April 4, 1978, defining
the Kentucky/Indiana border for purposes of decidingfrom which State the utility license applicant mustobtain a Section 401 water quality certificate for its
Marble Hill facility. On June 27, 1978, the D.C. Circuit dismissed the petition for review insofar as it
related to the licensing board decision, but retained
jurisdiction over the petition to the extent it sought

================================================271
review of the appeal board's February 16 decision.
This case has been argued and awaits decision.

State of New York and State of Illinois v. NRC, etal. (S.D.N.Y., 79 Civ. 4568).

A. R. Martin- Trigona v. Department of Justice, etal. (S.D. Ill., No. 78-4006).
On January 30, 1978, plaintiff sued the JusticeDepartment, Commonwealth Edison, and the NRC
concerning the withholding under the FOIA of
documents pertaining to the Quad Cities power station. NRC is asserting Exemption 7 as grounds forwithholding the documents. A motion to dismiss was
filed on April 21, 1978; the Court has not yet acted onthat motion.

State seeks to stop air shipment of plutonium, pending preparation of environmental impact statement.
In No. 79 Civ. 4568, the two States challenge the adequacy of the NRC's environmental statement ontransportation (NUREG·0170).

Coalition for the Environment, St. Louis Region
and Utility Consumers Council of Missouri v. NRC(D.C. Cir., No. 77 I905).
On October 5, 1977, petitioners sued to suspend theconstruction permit for the Callaway Nuclear Plant
based on a challenge to the Commission's interim fuelcycle rule. On December 1, 1977, the Court held thiscase in abeyance until 30 days after the Supreme
Court's decision in the Vermont Yankee case. On June7, 1978, the Court requested the parties' views as to
how this and the other fuel cycle cases should be
handled. NRC advised that it should be dismissed orheld in abeyance pending a final fuel cycle rule. The
case is still in abeyance, as of the close of the reportperiod.
w

w

Rosanna Kelly v. Hendrie, et al. (D.D.C., No.
79-1550) .
Plaintiff alleges that, in her efforts to be promoted,
she has suffered age and sex discrimination and that
she has experienced retaliation as a result of initiating
EEO procedures. Plaintiff seeks retroactive promotion
and an injunction against discrimination. NRC's
answer, filed in September 1979, denies the substantive allegations of her complaint.
State of New York v. NRC, et al. (S.D.N.Y., No. 75
Civ. 2121).
State of New York v. NRC, et al. (2d Cir., Nos.
75-6115,76-6022 and 76-6081).

Motion for preliminary injunction was denied in
September 1975. Appeal was taken to Second Circuit.
State's motion for summary judgment was filed on
December 11, 1975. The State's motion for summary
judgment was denied, and it sought review of thatdecision in the Second Circuit. The two appeals wereconsolidated, as was a third appeal taken from theDistrict Court's order dismiSSing the Civil Aeronautics
Board and the Customs Service as parties to the litiga
tion. Oral argument was heard on July 21, 1976. On
February 14, 1977, the Second Circuit issued a
39-page opinion affirming the District Court's denialof plaintiffs motion for a preliminary injunction. The
Court reasoned that plaintiff had failed to prove irreparable injury in view of the remoteness of atransportation accident and the absence of an agency
commitment of resources to a particular transportation mode. The Second Circuit also dismissed, on procedural grounds that a final appealable order was
lacking, the appeals from the District Court's dismissalof the CAB and Customs Service, and the denial ofplaintiffs summary judgment motion. The Second
Circuit noted NRC's expectation (stated in a
November 17, 1976 letter to the U.S. Attorney) that
the Commission's impact statement would be published early in February 1977. In fact, NUREG-0170 was
issued in December 1977. On September 6, 1978, NewYork filed its amended complaint, challenging theadequacy of the environmental statement. On October
27 1978 NRC filed the answer. With the filing of No.
79' Civ. '4568, these cases are being dismissed by
stipulation of the parties. NRC's answer to No. 79 Civ.
4568, along with interrogatories and a request for
documents, was filed on October 22, 1979.
M

14
Administration andManagement
NRC began setting up its operations andsupport headquarters at the TMI site assoon as the scope of the accident was
realized.

The NRC began fiscal year 1979 with an authorizedfull-time permanent personnel strength of nearly
2,800 and funding of $326 million. In response to theThree Mile Island accident, Congress approved in Julyan additional 100 positions and $5 million to be utilized by NRR, bringing authorized full-time permanent
personnel strength to 2,888 and funding to $331
million by the end of fiscal year 1979. NRC Headquarters activities remained dispersed in 10 buildings
throughout the District of Columbia and Maryland,with consolidation currently under review by the
Senate Committee on Environment and Public Works.
The Civil Service Reform Act of 1978, enacted October 13, 1978, has impacted employee and management practices in all Federal agencies. These and othermanagement and administrative support
developments, including organizational, personnel,
and fiscal matters are discussed below.

PERSONNEL AND ORGANIZATION
Augmented by an additional 100 positions from
Congress during the fiscal year, NRC's authorized fulltime permanent personnel strength had increased at
year-end to 2,888-more than 6 percent above thefiscal year 1978 level of 2,723. Over 70 percent of NRC
employees are located in the major program offices,
about 20 percent in program direction and administration, and the remainder are employed in the Commission, Commission staff, and the independent advisory
and adjudicatory bodies.
Of the 70 percent of NRC's employees holding college degrees, over 25 percent have masters degrees,
almost 4 percent have professional (mostly. law)
degrees, and more than 9 percent have doctorates.
Employees trained as scientists or engineers compriseover half of the NRC's work force.

Commission and Office Director
Appointments
Victor Gilinsky was reappointed by the President toserve a second term as a member of the Commission.
Mr. Gilinsky's reappointment is the first to the Commission, and extends from July 1, 1979 through June30, 1984. The Commission thus continued at full fivemember strength. On December 7, 1979, President
Carter designated Commissioner John F. Ahearne toserve as Chairman pending selection of a new Chairman from outside the agency. (See Chapters 1 and 2.)
In the upper management of the program offices,
Victor Stello, Jr., the former director of the Division of
Operating Reactors, Nuclear Reactor Regulation, wasnamed director of the Office of Inspection and Enforcement. This post had previously been filled byJohn G. Davis following the departure of Dr. Ernst
Volgenau. In the Commission staff offices, Leonard
Bickwit was named General Counsel, filling a post
which had been vacated by Jerome Nelson, and Albert
P. Kenneke was named acting director of the Office of
Policy Evaluation on the departure of Kenneth S.
Pedersen. In November 1979, Edward Hanrahanbecame director of the Office of Policy Evaluation.
Other key management changes included thefollowing three appointments: Max W. Carbon wasnamed chairman of the Advisory Committee on Reactor Safeguards; Robert M. Lazo, acting chairman of
the Atomic Safety and Licensing Board Panel; and E.
Kevin Cornell, Deputy Executive Director for Operations.

Organizational Changes
The organization and normal functions of many
NRC units were disrupted in the wake of the Three
Mile Island accident in March 1979 as resources werereallocated and priorities shifted, both in direct support of recovery efforts at the site and in related ac-274

tivities to pursue near-term and long-range objectives
whose needs were revealed by the event.
The organizational structures of the Offices of
Nuclear Reactor Regulation (NRR) and Inspection and
Enforcement (IE) were particularly affected as several
task groups launched activities on safety issues of thehighest priority. In May, NRR revamped the priorities
assigned to its activities and realigned its structure into
an interim organization designed for optimum use ofavailable resources. This required the temporary
reassignment of many personnel and a realignment ofmanagerial and technical talents which were expected
to continue well into 1980.
Staff from these and many other offices were usedfrom time to time at the Three Mile Island site-manyon a voluntary basis-for support during the first few
weeks following the accident. For most of the year, the
NRC Recovery Operations Office, manned bytechnical and some administrative staff, operated

from mobile office space at the TMI site. In
December, efforts were in progress to lease office spacein Middletown, Pa., for a projected long-term stay of
NRC staff in conjunction with decontamination effortsof the licensee at TMI Unit 2.
Other staff dislocations included the assignment of
55 personnel from throughout the NRC to support theagency's Three Mile Island Special Investigation
Group, formed in June, whose report to the Commission was expected to be issued in January 1980.
Reorganization Impending. President Carter, in his
December 7, 1979 statement on the Kemeny Commission report concerning the TMI accident, announced
he would send to Congress a reorganization plan tostrengthen the role of the NRC Chairman, and that a
new Chairman would be designated from outside theagency. In the meantime, he asked Commissioner
Ahearne to serve as the NRC Chairman. (See Chapters
1 and 2.) The President also directed the Federal

r========= 275

rights and employee protections.
the establishment of merit prinl:.er~;onnel practices and actions, anLabor Relations Authority, andService (SES).
the CSRA, with 98 percent
1I.o... ~· ........... to the SES system. The Ex, a management structure toprogram at the NRC, was
II;OInmlissllon in August 1979. It is
Directors, is chaired by the Extions, and is organized intog Committee, the Performance
tive Resources Management".Li.n.'V....... tive Development CommitCSRA, such as the merit pay
and probationary periods for, were not required of the
excepted status of its personnel
is, however, undertaking studies
to participate in these and othertation Election. In late 1978, themnIIO'\l'ee~ Union (NTEU) petition-of Labor seeking to represent
in the five NRC regional ofelection resulted in the certifica-

276

tion of the NTEU, on January 26, 1979, as the exclusive representative of both headquarters andregional office employees in one bargaining unit.
Subsequent Negotiations. Soon after certification,
the NTEU proposed the negotiation of an interim procedure for resolving employee grievances andgrievances which may be brought by the union or by
NRC management as an institution. These negotiations began on February 27, 1979 and were concluded on June 6, 1979, with full accord on an interim
agreement governing the disposition of grievances andprocedures for invoking arbitration.
On June 27, 1979, the NTEU forwarded proposals
to NRC for a comprehensive bargaining agreement
covering the full range of personnel policies and practices. These proposals were under study at year-end,
with negotiations anticipated early in fiscal year 1980.

Equal Employment Opportunity
Progress at GS-ll and Up. In fiscal year 1979, the
NRC continued to submit quarterly reports to theCongress on minorities and women hired and promoted, and on other actions to improve the agency's
EEO posture, with main emphasis on grade levels
GS-ll and above. Of all new employees hired at
GS-ll and higher grades during fiscal year 1979, 11.5
percent were from minority groups and 11.5 percent
were women. Promotions to grades GS-l1 and above
numbered 374, with minority personnel representing
10.2 percent and women constituting 17.6 percent. On
September 30, 1978, there were 126 minority and 143
women employees at GS-11 and above. By September
30, 1979, these figures had changed to 174 and 178,
respecti vel y .
Recruiting Emphasis. Recruiters visited 25 colleges
and universities during the year to attract candidates
for the NRC Intern, Cooperative Education, and Summer Intern Programs. Visits to nine of these schools

were for the primary purpose of recruiting minority
and women employees. Of the 145 persons hired underthese programs, 81 were minority personnel and/or
women. Continued EEO emphasis is expected tosignificantly increase the future representation ofminorities and women in the higher grades. .
Special EEO Programs. The special EEO programs
highlighting Hispanic heritage, Black history,
awareness training, and career counseling continued
during fiscal year 1979. In addition, the NRC observed
the first National Asian/Pacific Heritage Week withactivities featuring contributions of Asian/Pacific
Americans to the development of the United States.
Women's Programs. Throughout fiscal year 1979,
the Federal Women's Program Advisory Committee
continued meetings with FWP members and key officedirectors to discuss questions raised by female
employees. FWP managers were appointed to each of
the NRC's five regional offices. The F'WP also conducted eight weekly workshops on "Leadership for
Women," which examined the various techniques,characteristics, and problems of leadership and success.
In June 1979, the NRC's FWP headquarters
manager was elected president of the Suburban
Maryland Chapter/Federally Employed Women,
Inc., while the legal advisor to the F'WPAC received
FEW's Outstanding Achievement Award. FWP
members also participated in the United Nations MidDecade Conference for Women, the first Minority
Women's Conference, the Image Convention, theWomen's Network Reception for U.S. Congresswomen, the Tenth Anniversary of FEW, Inc"
receptions for the Women's Airforce Service Pilots,
salute to Asian/Pacific American Women, and conferences for FEW and the Society for WomenEngineers.
Class Actions. Four class action complaints are
pending against the NRC alleging sex discrimination.
An Equal Employment Opportunity Commission

The Howard University Jazz Ensemble played to a good "house"
during noon-hour performances for members of the NRC staff.
The group was featured as part of NRC's Black History observancein 1979.

277(EEOC) complaints examiner has been appointed in
the first case to conduct a hearing on the issue of
whether NRC's policy of auditing positions in responseto requests for promotions has had a discriminatory
impact on females. The three remaining complaints
are pending EEOC's recommendations regarding procedural sufficiency and a determination as to whether
the four cases should be consolidated.

INSPECTION AND AUDITThe Office of Inspector and Auditor (IA) conducts
audits, investigations and inspections to assure the effectiveness, efficiency and integrity of NRC operations. Its responsibilities have included reviews of
employee complaints and financial, compliance andmanagement audits, as well as liaison functions with
the General Accounting Office and Department ofJustice. Some of the more important IA activities are
summarized below.

Procurement of Goods and Services
At the request of the Executive Director for Operations, IA performed an inspection to determine
whether NRC's procurements of goods and services
were accomplished properly and whether such goods
and services were fully received. This review was both
timely and significant in light of the recent investigations of procurement abuses in the Federal government. lA's report of June 1979 evaluated internal con·troIs and procedures, reviewed the certification of
payment processes, and commented on the custody
and accountability of certain goods ordered andreceived. Based on selected tests and observations, IA
determined that most goods were received and services
rendered. However, some discrepancies existed. The
IA report pointed out a need for better planning to
avoid waste. IA believes that the EDO's proposed corrective actions should resolve the problems noted in
the report.

Resident Inspection Program
The accident at Three Mile Island (TMI) altered theoriginal scope of this review to include assessment ofproblems and issues which might occur if implementation of the Resident Inspection Program at reactor
facilities was accelerated. The report, issued in June
1979, concludes that accelerating this program would
be difficult and would reduce the scope and quality of
the inspection effort now being performed.

Guidance for Safeguards- Upgrade Rule
A report issued in December 1979 reviews the planning and development of guidance to licensees in
response to the July 24, 1979, proposed rule to upgrade
physical security at fuel cycle facilities. In December
1976, the Commission endorsed a general upgrading
based on the recommendations of a Joint ERDA/NRC
Task Force; since January 1977, the safeguards staffs
have been in the process of developing the rule and theguidance necessary to implement it.
The IA report discloses that a fragmented approachto the planning and development of safeguards intormation has resulted in nearly $400,000 worth oftechnical reports that will not be used as gUidance,
and in the use of outside assistance for the development of information that could have been done inhouse. The report also identifies the NRC's need to improve safeguards interagency coordination to help prevent possible duplication of safeguards information inareas where other agencies have parallel interests andsimilar ongoing projects. It further criticizes theusefulness of certain documents in a draft compendium of existing NRC and non-agency guidance
related to the upgrade rule. The compendium was firstreleased in March 1979, and contains language that
may be confusing or misleading to licensees, as well as
guidance documents and evaluation tools which may
be crucial to licensees in upgrading their physicalsecurity systems, but which may not be available during the rule's required implementation period.

Automatic Data Processing
Two reports were issued which review accountability of automatic data processing equipment andrequirements, and management of NRC's ADP
systems. In March 1979, IA issued a report assessing
NRC's ability to determine long-range ADP requirements. This contained recommendations tostrengthen controls over the procurement of and accountability for ADP equipment. The second report,
issued in August 1979, concentrated on NRC's
management and organizational structure for ADP. As
in the first report, IA made recommendations for amore economical and efficient operation.

Review of Semiscale Program
A report on lA's review of the Semiscale research
program was issued in September 1979. The review
confirms that this program and related activities arebeing carried out in an efficient and effective manner;278

however, it also reveals a need for improvement in theareas of analysis of direct and indirect costs and in thedissemination and use of Semiscale technical reports.
The report makes appropriate recommendations.

Levels of Review of a Significant Report
IA conducted an investigation to chronologicallyreconstruct, up until the March 28, 1979 accident at
TMI, the events and levels of review in the analysis of
the report, "Decay Heat Removal During a Very Small
Break LOCA for a B&W 205 Fuel-Assembly PWR."
This report, which was prepared in 1977 by an
employee of the Tennessee Valley Authority, raised
concerns of apparent relevance in the wake of the accident at TMI. (See also Chapter 2.) lA's investigation
identified the extent of the NRC review process, andthe rationale for certain actions taken by cognizant individuals in the chain of review.

FUNDING AND BUDGET MATTERS
NRC resource charts and financial statements appear at the end of this chapter. These charts show
allocations of authorized personnel and funds to thevarious NRC activities carried out during fiscal year
1979, and to those projected for fiscal year 1980.
The increase in personnel for fiscal year 1980 is
mainly in support of the expanded resident inspector
program, and to provide an additional staff of 100 to
NRR for non-TMI related licensing activities. These
positions were allocated by Congress after budget approval, since the TMI accident had not yet occurred.
The increase in funding for fiscal year 1980 results
primarily from emergency core cooling system experiments for light water reactors requiring increased
funds to maintain the research efforts initiated in prioryears and the first full year of LOFT nuclear testing.
Also, increased personnel compensation funding is required to support the added personnel noted earlier.
The financial statements following the charts are
self-explanatory.

Contracting and Reimbursable Work
Most of the NRC's operating funds are expended inreimbursable arrangements with other agencies andcontracts for confirmatory research and technicalassistance in every major area of the agency's activity.
Approximately $196 million was allocated to program support during fiscal year 1979, of which $171
million went for reimbursable work performed for the
NRC by other Federal agencies. The Department ofEnergy's share of this was approximately $169 million
for work performed in DOE's national laboratories
and other facilities. This work included major research
projects and experiments at the Loss-of-Fluid Test

(LOFT) Facility, the Power Burst Facility, and the
Semiscale Facility. (Specific research programs aredescribed in Chapter 11.)
Technical and administrative assistance contracts(except work performed through DOE), as well as
general purchases of all kinds, are administeredthrough the Division of Contracts in the Office of Administration. Such contracts totaled more than $44
million during fiscal year 1979.
Major activities in the Division of Contracts to improve procurement practices during the fiscal year
have focused on (1) publication of a final rule on the
avoidance of organizational conflicts of interest; (2)promulgation of an NRC manual chapter on the treatment of unsolicited proposals in the agency; (3)development of internal procedures for the procurement of goods and services; (4) implementation of abidder's mailing list system for firms who desire to furnish goods and services to the NRC; and (5) development of a management information system which will
provide the data required by the Federal Procurement
Data System.
Additionally, source selection criteria were formalized for determining the most appropriate source
for fulfilling NRC requirements for goods and services.

Representatives of TERA Advanced Services Corporation, contractor for the installation and operation of NRC's automated document control system, spent many hours in 1979 working with NRC
personnel on use of the system. This photo shows TERA official
Jim Long conducting a typical DCS training session.

AUTOMATED CONTROL OF
DOCUMENTS
The NRC's computerized Document Control System(DCS) is operated under contract, and is used to index,
store, and retrieve NRC documents. Each document
received or generated by the NRC is indexed bybibliographic elements and microfiched for storage
and remote access. This allows up to 12 index searches

===================================================279
of the data base by computer and production of page
images on remote video screens. The DCS is also providing periodic indexes and title list reports, including
a daily accession list of documents placed in the PublicDocument Room, that were formerly produced underhigh cost time-sharing contracts.
The monthly Title List of Documents Made Publicly
Available was first published in January 1979. ThisDCS computer-produced document replaces Power
Reactor Docket Information (PRDI), which was issued
through DOE's Technical Information Center, andcontained only selected docketed information. The
final issue of PRDI, also published in January 1979,
was an annual cumulation of the information for
1978. The new publication includes docketed material
associated with civilian nuclear power plants andother uses of radioactive material, as well as nondocketed material received and generated by the NRC
pertinent to its regulatory role.
Following the Three Mile Island (TMI) accident,
the contractor processed on an accelerated schedule all
pre- and post-incident TMI-2 documents held by the
NRC, and produced a titled microfiche for each. The
DCS was then used to print complete title listings onrequest, cumulative through any date, to assist in theongoing reviews of the accident. The search, indexing,
and microfiche features of the DCS contributed
significantly to the stafFs efforts following the accident, and prompted a rescheduling of the major contract tasks to permit an accelerated backfit of otherpower plant docket files onto the data base.
The DCS facility is located near NRC headquarters
in Bethesda, Md., and houses the contractor's staff ofengineers, computer specialists, indexers, technicalcoders, and the computer and microfiche equipment.
Approximately 1,000 documents per day are being
processed through the facility.

PHYSICAL FACILITIES
During fiscal year 1979, the NRC continued tohouse approximately 2,400 headquarters employees in
10 buildings-one located in the District of Columbia,
and nine in the Maryland suburbs.
In October 1977, the House Committee on Public
Works and Transportation approved a General Services Administration (GSA) report which proposed
consolidation of NRC headquarters in a single facility
to be constructed on an urban renewal site in
Washington, D.C. (See 1977 NRC Annual Report,
page 208.) However, as a result of differing views
presented by employee representatives and Maryland
officials and legislators at an April 1978 hearing of the
Subcommittee on Public Buildings and Grounds of the
Senate Committee on Environment and Public Works,that committee requested GSA to analyze several additional factors. One of the main factors to be considered
was the feasibility of consolidating in Montgomery
County, Maryland.

GSA's expanded study, submitted to the Senate
committee on July 25, 1978, recommended that two
locations in the District of Columbia and three sites inMontgomery County be considered. Prior to final action, however, the committee imposed a moratorium
on approval of all pending leasing and construction activities. The moratorium had not been lifted by theend of the fiscal year.
In late 1979, GSA began amending the originalreport to Congress in order to update the factual information made obsolete by increased costs and by the
NRC's expanded responsibilities resulting from theThree Mile Island accident. Congressional consideration of lifting the construction moratorium, and particular attention to alleviating NRC's headquarters
dispersion, is expected.

NRC LICENSE FEESOn March 23, 1978, the Commission adopted arevised schedule of license fees. (See 1978 Annual
Report, pp. 253-255.) The schedule increased fees inseveral categories of applications and licenses, andestablished additional categories of cost recovery forgovernment services. These new categories included
inspections, amendments, applications filed by vendors and architect-engineers for approval of standardized designs, and renewals.
The NRC expects to recover less than 3 percent of its
budget through the increased fees. Those collected infiscal year 1979 amounted to $12.5 million, of which$2.6 million is held in a suspense account by the
Department of Treasury until calculations of actual
costs after action on the permit or license involved
have been completed. The total collected since fees
were first imposed in 1968, through September 1979,
was $101.4 million, of which $6.5 million has beenrefunded.
On February 5, 1979, the United States Court ofAppeals for the Fifth Circuit heard oral arguments by
17 utilities and 2 waste disposal licensees challenging
the NRC's interpretation of the Independent Offices
Appropriation Act of 1952, which is the statutory
authority for the Commission schedule of fees in 10
CFR Part 170.
On August 24, 1979, the Fifth Circuit issued its
opinion upholding in all respects the Commission's
February 9, 1978 license fee schedule and its
guidelines for fees. The Court specifically held that the
NRC: (1) has the authority to recover the full cost ofproviding services to identifiable beneficiaries; (2)may recover the costs it incurs in conducting routine
inspections, complying with the National Environmental Policy Act, conducting uncontested hearings, and providing administrative and technical support services; and (3) may assess fees to recover thecosts of services it provides to Agreement State
licensees who may also require an NRC license for certain activities.

280

NATIONAL EMERGENCY
PREPAREDNESS
During early 1979, efforts were continued toward
insuring NRC capabilities in the event of a national
emergency, including an attack on the United States.
The accident at Three Mile Island, however, interrupted the work on national emergency preparedness
as staff efforts were increased to assist State and local
governments in their radiological emergency
preparedness.
An agreement was reached to permit NRC to sharethe use of the Department of Housing and Urban

Development's alternate emergency operating facility.
This facility would be the place of operations for the
NRC "Charlie" team, if the NRC management
capability needed to be dispersed in an emergency.
After the Federal Emergency Management Agency
was formed in July 1979, contacts and lines of communication were confirmed with the new agency forcontinued cooperation in national emergency
preparedness.

NRC RESOURCESFY 1979

ProgramDirection &Administration

Materiel
Safety &Safeguards

ACRS.
Boards &Legal

PERSONNEL - 2691

FUNDS - $326 MILLION

NRC RESOURCES
FY1980ProgramDirection &Administration

Material
Safety &Safeguards

ACRS.
Boards &Legal

PERSONNEL -3034

FUNDS - $369 MILLION

281

Fiscal Year 1979-NRC Financial Statements
Balance Sheet (in thousands)
AssetsCash:Appropriated Funds in U.S. Treasury
Other (See Notes 1 and 3)

September 30,
1979

September 30,
1978$ 146,257
10,426
156,683$ 129,149
10,841
139,990

222
5,986
272
6,480

142
3,259
85
3,486

7,462
1,314
6,148

5,716
1,428
4,288

171
1,304
1,475

129
1,113

$ 170,786$ 148,877

Accounts Receivable:
Federal Agencies
Miscellaneous Receipts (Note 2)
Other

Plant:
Completed Plant and Equipment (Note 5)
Less-Accumulated Depreciation (Note 5)

Advances and Prepayments:
Federal Agencies
Other

Total Assets

Liabilities and NRC Equity
Liabilities:
Funds held for Others (See Notes 1 and 3)
Accounts Payable and Accrued Expenses:
Federal Agencies
Other
Accrued Annual Leave of NRC Employees
Deferred Revenue (Note 3)
Total Liabilities
NRC Equity: Balance @ October 1, 1978
Additions:
Funds Appropriated-net
Deductions:
Net Cost of Operations (Note 5)
Funds Returned to U.S. Treasury (Note 2)
Total NRC Equity
Total Liabilities and NRC Equity

Septem ber 30,
1979$

10,426

September 30,
1978$

10,84142,884
20,3236,285
1,330
81,24875,924

39,17615,3185,552
2,067
72,954
68,914

326,601
402,525

290,023
358,937

305,865
7,122
312,987
89,648273,153
9,861
283,01475,923$ 170,786

Note 1. As of September 30, 1979, includes $2,504,163.30 of funds received under cooperative research agreements involVing NRC, DOE,
Federal Republic of Germany, Japan, Austria, and the Netherlands. Included also is $7,117,130.00 of funds received from deferred
revenue billings. These funds will be refunded and/or recorded as earned revenue after the cost of processsing the applicable application
has been finalized and, accordingly, are not available for NRC use. (See Note 3.)
Note 2. These funds are not available for NRC use.
Note 3. On March 24, 1978, 10 CFR Part 1 was revised. Contained therein by category of license are maximum fee amounts to be paid by applicants at the time a facility or material license is issued. Also, after the review of the license application is complete (generally after
license -has been issued), the expenditures for professional manpower and appropriate support services are to be determined and theresultant fee assessed. In no event will the fee exceed the maximum fee for that license category which generally has been paid. This
could involve the refunding of a significant portion of the initial amount paid. Therefore, the revenue is recorded in a deferred revenue
account at the time of billing and is removed from this account and recorded in Funds Held for Others when the bill is paid. The balancein the Deferred Revenue account consists of deferred revenue on billings issued but not collected. (See Note 1.)

282====================~============================

Fiscal Year 1978/1979 Statement of Operations (in thousands)
Fiscal Year 1979
Fiscal Year 1978
(October 1, 1978, (October 1, 1977,
thru September 30, thru September 30,
1979)
1978)
Personnel Compensation
Personnel Benefits
Program Support
Administrative Support
Travel of Persons
Training (Technical)
Equipment (Technical) (See Note 4)
Construction (See Note 4)
Taxes and Indemnities
Refunds to Licensees
Representational Funds
Reimbursable Work
Increase in Annual Leave AccrualDepreciation Expense
Equipment Write-offs and Adjustments
Total Cost of Operations
Less Revenues:
Reimbursable Work for Other Federal Agencies
Fees (deposited in U.S. Treasury as Miscellaneous
Receipts (See Note 2))
Indemnity
Material Licenses
Facility Licenses
Other$ 85,144
7,649
181,950
27,325
6,123
585
6,545
10
3
180
9
367
733
547
26

$

77,144
7,172
161,817
19,120
5,378
759
7,687
1,672
5
189
9
273
709
469
229

$ 317,403$

282,632

367

273

1,035
1,605
7,810
137

1,793
321
7,383
1,188

306,449

271,674

584

1,479

Total Revenue
Net Cost of Operations before Prior Year Adjustment
Prior Year Adjustment (See Note 5)
Net Cost of Operations$ 305,865

u.S. Government Investment In The Nuclear Regulatory Commission(From January 19, 1975, Through September 30, 1979-in thousands)
Appropriation Expenditures:
Fiscal Year 1975 (January 19, 1975, through June 30, 1975)
Fiscal Year 1976 (July 1, 1975, through September 30, 1976)
Fiscal Year 1977 (October 1, 1976, through September 30, 1977)
Fiscal Year 1978 (October 1, 1977, through September 30, 1978)
Fiscal Year 1979 (October 1, 1978, through September 30, 1979)$

52,792
226,248
230,559
270,877$ 309,493
1,089,969

Unexpended Balance of Appropriated Funds in U.S. Treasury, September 30, 1978

146,257

Transfer of Refunds Receivable from Atomic Energy Commission, January 19, 1975

429

Total Funds AppropriatedLess:
Funds returned to u.s. Treasury (See Note 2)
Assets and Liabilities transferred from Other Federal Agencies without Reimbursement
Net Cost of Operations from January 19, 1975, through September 30, 1979

$1,236,655
50,741
2,018
1,094,358

Total Deductions
NRC Equity at September 30, 1979, as shown on Balance Sheet
Note 4. Represents current year cost of plan and equipment acquisitions for use at DOE facilities.
Note 5. During fiscal year 1979, net cost of operations was reduced by $584,478.98 for net write-ons of capital equipment resulting from thereconciliation of the November 1977 physical inventory.

================================================283
Appendix 1

NRC ORGANIZATION(As of September 30, 1979)
COMMISSIONERSJoseph M. Hendrie, Chairman'"
Victor Gilinsky
Richard T. KennedyPeter A. Bradford
John F. Ahearne
The Commission Staff
General Counsel, Leonard Bickwit
Office of Policy Evaluation, Albert P. Kenneke, Acting Director'" '"Office of Public Affairs, Joseph J. Fouchard, Director
Office of Congressional Affairs, Carlton C. Kammerer, Director
Office of Inspector and Auditor, James J. Cummings, Director
Secretary of the Commission, Samuel J. Chilk
Other Offices
Advisory Committee on Reactor Safeguards, Max W. Carbon, Chairman
Atomic Safety & Licensing Board Panel, Robert M. Lazo, Acting Chairman
Atomic Safety & Licensing Appeal Panel, Alan S. Rosenthal, Chairman
EXECUTIVE DIRECTOR FOR OPERATIONS
Executive Director for Operations, Lee V. Gossick
Deputy Executive Director for Operations, E. Kevin Cornell
Technical Advisor, VacantProgram Offices
Office of Nuclear Reactor Regulation, Harold R. Denton, Director
Office of Nuclear Material Safety and Safeguards, William J. Dircks, Director
Office of Nuclear Regulatory Research, Saul Levine, Director
Office of Standards Development, Robert B. Minogue, Director
Office of Inspection and Enforcement, Victor Stello, Jr., Director
Staff Offices
Office of Administration, Daniel J. Donoghue, Director
Executive Legal Director, Howard K. Shapar
Controller, Learned W. Barry
Office of Equal Employment Opportunity, Edward E. Tucker, Director
Office of Management and Program Analysis, Norman M. Haller, Director
Office of International Programs, James R. Shea, Director
Office of State Programs, Robert G. Ryan, Director
Office for Analysis and Evaluation of Operational Data, C. J. Heltemes, Interim Director
Regional Offices
Region I Philadelphia, Pa., Boyce H. Grier, Director
Region II Atlanta, Ga., James P. O'Reilly, Director
Region III Chicago, Ill., James G. Keppler, Director
Region IV Dallas, Texas, Karl V. Seyfrit, Director
Region V San Francisco, Calif., Robert H. Engelken, Director"'On December 7, 1979, President Carter designated Commissioner Ahearne to serve as Chairman.
'" "'Edward Hanrahan was named Director of PE in November 1979.

284

The NRC is responsible for licensing and regulating
nuclear facilities and materials and for conducting researchin support of the licensing and regulatory process, as mandated by the Atomic Energy Act of 1954, as amended, theEnergy Reorganization Act of 1974, as amended, and the
Nuclear Non-proliferation Act of 1978; and in accordancewith the National Environmental Policy Act of 1969, as
amended, and other applicable statutes. These responsibilities include protecting public health and safety, protecting the environment, protecting and safeguarding materials
and plants in the interest of national security; and assuringconformity with antitrust laws. Agency functions are performed through: standards-setting and rulemaking; technicalreviews and studies; conduct of public hearings; issuance ofauthorizations, permits and licenses; inspection, investigationand enforcement; evaluation of operating experience, andconfirmatory research. The Commission itself is composed offive members, appointed by the President and confirmed by
the Senate, one of whom is designated by the President as
Chairman. The Chairman is the principal executive officer
and the official spokesman of the Commission.
The Executive Director for Operations directs and coordinates the Commission's operational and administrative activities and the development of policy options for Commission
considera tion.
The Office of Nuclear Reactor Regulation licenses nuclear
power, test and research reactors under a two phase process.
A construction permit is granted before facility construction
can begin and an operating license is issued before fuel can beloaded. NRR reviews license applications to assure that the
proposed facility can be built and operated without undue
risk to the health and safety of the public and with minimal
impact on the environment. NRR monitors operating reactor
facilities during their lifetime through decommissioning.
NRR also reviews the financial responsibility of each applicant for a construction permit, confirms that each applicant
is properly indemnified against accidents, and verifies thatthe applicant(s) is not in violation of antitrust laws.
The Office of Nuclear Material Safety and Safeguards is
responsible for ensuring public health and safety, and protection of national security and environmental values in thelicensing and regulation of facilities and materials associatedwith the processing, transport, and handling of nuclear
materials. NMSS reviews and assesses safeguards against
potential threats, thefts, and sabotage, and works closelywith other NRC organizations in coordinating safety andsafeguards programs and in recommending research, standards, and policy options necessary for their successful operation.
The Office of Nuclear Regulatory Research plans and implements research programs of nuclear regulatory researchwhich are deemed necessary for the performance of the Commission's licensing and regulatory functions. Research programs cover reactor safety areas such as materials behavior,site safety, systems engineering, and computer code development and assessment. Research is also performed on
safeguards, health effects associated with the nuclear fuel cycle, environmental impact of nuclear power, waste treatment
and disposal, and transportation of radioactive materials.
The Office of Standards Development develops regulations, guides, and other standards needed for regulation offacilities and materials with respect to radiological healthand safety and environmental protection, for materials
safeguards and plant protection, and for antitrust review.
The Office also coordinates NRC participation in national
and international standards activities.
The Office of Inspection and Enforcement inspects nuclear
facilities and materials licensees to determine whether
facilities are constructed and operations are conducted incompliance with license provisions and Commission regulations, and to identify conditions that may adversely affect theprotection of nuclear materials and facilities, the environment, or the health and safety of the public; inspects applicants and their facilities to provide a basis for recommending issuance or denial of licenses; investigates accidents, incidents, and allegations of improper actions that involve
nuclear material and facilities; and enforces NRC regulations
and license provisions. IE, on behalf of NRC, manages anddirects the Commission's five regional offices, located in
Philadelphia, Pa., Atlanta, Ga., Chicago, Ill., Dallas, Texas,
and San Francisco, Calif.
THE COMMISSION STAFF
The Office of the Secretary provides secretariat services forthe conduct of Commission business and implementation ofdecisions, including planning meetings and recording
deliberations, manages the staff paper system, monitors thestatus of actions, and maintains the Commission's official
records. The office also processes institutional correspondence, controls the service of documents in adjudicatory and public proceedings, supervises the
Washington, D.C. Public Document Room, administers the
NRC historical program, and provides administrative support for the Commission.
The Office of General Counsel serves the Commission in avariety of legal capacities. The Office assists the Commission
in the review of Appeal Board decisions, petitions seeking
direct Commission relief, and rulemaking proceedings, anddrafts legal documents necessary to carry out the Commission's decisions. The General Counsel provides a legal
analysis of proposed legislation affecting the Commission'sfunctions and assists in drafting legislation and preparing
testimony. The General Counsel also represents the Commission in court proceedings, frequently in conjunction with the
Department of Justice.
The Office of Policy Evaluation plans and manages activities involved in performance of an independent review ofpositions developed by the NRC staff which require policy
determinations by the Commission. The Office also conducts
analyses and projects which are either self-generated or requested by the Commission.
The Office of the Inspector and Auditor investigates toascertain the integrity of all NRC operations; investigatesallegations of NRC employee misconduct, equal employment
and civil rights complaints, and claims for personal property
loss or damage; conducts the NRC's internal audit activities;
and hears individual employee concerns regarding Commission activities under the agency's "Open Door" policy. The
office develops policies governing the Commission's financial
and management audit program and is the agency contact
with the General Accounting Office on this function. Refers
criminal matters to the Department of Justice and maintains
liaison with law enforcement agencies.

285

The Office of Public Affairs plans and administers NRC's
program to inform the public of Commission policies, pro·grams and activities and keeps NRC management informedof public affairs activities of interest to the Commission.
The Office of Congressional Affairs provides advice andassistance to the Commission and senior staff on congressional
matters, coordinates NRC's congressional relations activities,
and maintains liaison for the Commission with congressional
committees and members of Congress.
SUPPORT STAFF
The Office of Administration directs the agency's programsfor organization and personnel management; security andclassification; technical information and document control;
facilities and materials license fees; contracting and procurement; rules, proceedings and document services; data processing; management development and training; and otheradministrative housekeeping and special services.
The Office of the Controller develops and maintains theCommission's financial management program, including accounting, budgeting, pricing, contract finance, automatic
data processing equipment acquisition, and accounting for
capitalized property. Prepares reports necessary to the
management of NRC funds. Maintains liaison with the
General Accounting Office, Office of Management andBudget, Congressional Committees, other agencies, and in·
dustry. The Controller also prepares the NRC Five-Year Planand performance resource evaluation studies.
The Office of the Executive Legal Director provides legal
advice and services to the Executive Director for Operations
and staff, including representation in administrative proceedings involving the licensing of nuclear facilities andmaterials, and the enforcement of license conditions andregulations; counseling with respect to safeguards matters,
contracts, security, patents, administration, research, personnel, and the development of regulations to implement applicable Federal statutes.
The Office of Equal Employment Opportunity develops
and recommends overall policy providing for equal employment opportuqity, recommends improvements or corrections
to achieve this goal, and monitors the agency's affirmative action program.
The Office of International Programs plans and implements programs of international nuclear safety cooperation, creating and maintaining relationships with foreign
regulatory agenCies and international organizations; coordinates NRC export-import and international safeguards
policies; issues export and import licenses; and coordinates
responses by NRC to other agencies related to export-importactions and issues.
The Office of Management and Program Analysis provides
NRC staff with management information and program
analyses; identifies and analyzes major NRC policy, programand management issues and conducts long- and short-range

planning to assist NRC operating officials; develops and implements management information and control systems andrecommends policy on use of such systems for agency-wide
applications; develops and implements application of soundstatistical practices within NRC; and coordinates special information projects on overall NRC policies and programs.
The Office of State Programs directs programs relating toregulatory relationships with State governments andorganizations and interstate bodies; manages the NRC State
Agreements program; and provides Federal agency leadership in assisting State and local governments in radiologicalemergency response planning.
The Office for Analysis and Evaluation of Operational
Data assures the proper analysis of operational dataassociated with all NRC-licensed activities and the feedbackof such analyses to improve safety. The office identifies key
analyses to be conducted, taking into account such factors as
postulated accident sequences and data availability; selects
appropriate analytical techniques and proposes data gathering mechanisms for data not currently available; conducts
systematic safety analyses and evaluations of operational data
to seek trends that would forecast a potential problem;
develops recommendations to resolve problems revealed byoperational data analyses and evaluations; provides
analytical guidance to, accepts technical input from, andcoordinates efforts of operational data analysis groups inother NRC offices; reviews overall NRC and industry
response to assess implementation of recommended actions;and serves as focal point for interaction with the ACRS andindustry groups involved in operational data analysis andevaluation.

OTHER OFFICES
Advisory Committee on Reactor Safeguards. A statutory
committee of 15 scientists and engineers, advises the Commission on the safety aspects of proposed and existing nuclear
facilities and the adequacy of proposed reactor safety standards, and performs such other duties as the Commission may
request.
Atomic Safety and Licensing Board Panel. Three-member
licensing boards drawn from the Panel-made up of lawyers
and others with expertise in various technical fields-conduct
public hearings and make such intermediate or final decisions
as the Commission may authorize in proceedings to grant,suspend, revoke, or amend NRC Licenses.
Atomic Safety and Licensing Appeal Panel.
Threemember appeal boards selected from the Panel exercise theauthority and perform the review functions which would
otherwise be carried out by the Commission in licensing proceedings. ASLB decisions are reviewable by an appeal board,either in response to an appeal or on its own initiative. The
appeal board's decision also is subject to review by the Commission on its initiative or in response to a petition for discretionary review.

i

I

l

.

,

...

Appendix 2

NRC Committees and Boards'eactor Safeguards

Atomic Safety and Licensing Board Panelstatutory committee in 1957 by Secergy Act of 1954, as amended. The
studies and facility license applicardance with the Atomic Energy Act
ation Act and makes reports thereon
he public record of the proceeding.
dvice with respect to the hazards ofcilities and the adequacy of related
mittee also performs such other addiission may request. The members areerms by the Commission. The comwn chairman and vice chairman. Ase members were:

Section 191 of the Atomic Energy Act of 1954 authorizes
the Commission to establish one or more atomic safety andlicensing boards, each comprised of three members, one ofwhom is to be qualified in the conduct of administrative proceedings and two of whom will have such technical or otherqualifications as the Commission deems appropriate to theissues to be decided. The boards conduct such hearings as theCommission may direct and make such intermediate or final
decisions as it may authorize in proceedings with respect togranting, suspending, revoking, or amending licenses orauthorizations. The Atomic Safety and Licensing Board
Panel (ASLBP) Office-with a permanent chairman who
coordinates and supervises the ASLBP activities-serves as
spokesman for the panel, and makes policy recommendations
to the Commission concerning conduct of hearings and hearing procedures. Pursuant to subsection 201 (g)(l) of theEnergy Reorganization Act of 1974, the functions performedby the licensing boards were specifically transferred to the
Nuclear Regulatory Commission. As of September 30, 1979
the ASLBP was composed of the following members and professional staff (""''' denotes full-time ASLBP members andstaff) :Chairman, Professor and Chairman, Department, University of Wisconr, Engineering Division, Oak Ridge
ak Ridge, Tenn.
ed Head Nuclear Engineer, Division
Tennessee Valley Authority, Knox'ON, Consulting Engineer (Mechanig), Jupiter, Fla.
, Professor of Nuclear Engineering,
emorial-Phoenix Project, University
r, Mich.
lOSKI, Senior Engineer, Chemical, Argonne National Laboratory,
IS, Department of Physics, UniversiBarbara, Calif.
Retired Division Leader, Los Alamos
Los Alamos, N. M.
" Retired Director, Planning, United'., Richland, Wash.
__ ER, Chairman, Department of Enciences, School of Public Health,oston, Mass.
Professor, School of Engineering andversity of California, Los Angeles,
ET, Professor, Department of Engieritus, California Institute of, Calif.
Retired Chief Electrical Engineer,
ompany, Philadelphia, Pa.
ON, Professor, Chairman of MetalDepartment, Ohio State University,
S, Professor, Head of Civil Engineerrersity of Illinois, Urbana, Ill.

ROBERT M. LAZO, Acting Chairman, ASLBP Attorney,
U.S. Nuclear Regulatory Commission, Bethesda, Md."
DR. GEORGE C. ANDERSON, Department of
Oceanography, University of Washington, Seattle, Wash.
CHARLES BECHHOEFER, ASLAB Attorney, Bethesda,
Md."
ELIZABETH S. BOWERS, ASLBP Attorney, Bethesda,
Md.'"
JOHN H. BREBBIA, Attorney with law firm of Alston,
Miller & Gaines, Washington, D.C.
GLENN O. BRIGHT, ASLBP Engineer, Bethesda, Md.
DR. A. DIXON CALLIHAN, Retired Physicist, Union Carbide Corporation, Oak Ridge, Tenn.
DR. E. LEONARD CHEATUM, Retired Director of Institute of Natural Resources, University of Georgia,
Watkinsville, Ga.
HUGH K. CLARK, Retired Attorney, E. I. duPont deNemours & Company, Kennedyville, Md.
DR. RICHARD F. COLE, ASLBP Environmental Scientist,
Bethesda, Md."
DR. FREDERICK P . COWAN, Retired Physicist,
Brookhaven National Laboratory, Stuart, Fla.
VALENTINE B. DEALE, Attorney at Law, Washington,
D.C.
RALPH S. DECKER, Retired Engineer, U.S. Atomic Energy
Commission, Cambridge, Md.
DR. DONALD P. DE SYLVA, Professor, Biology and Living
Resources, School of Marine and Atmospheric Science,
University of Miami, Miami, Fla.
>I<

287

MICHAEL A. DUGGAN, College of Business Administra~tion, University of Texas, Austin, Tex.
DR. GEORGE A. FERGUSON, Professor of Nuclear
Engineering, Howard University, Washington, D.C.
DR. HARRY FOREMEN, Director, Center tor .PopUlatIOn
Studies, University of Minnesota, Minneapolis, Minn.
JOHN H. FRYE, III, ASLBP Legal Counsel, Bethesda,
Md.'"
MICHAEL GLASER, Partner, law firm of Glaser and Flet~cher, Washington, D.C.
ANDREW C. GOODHOPE, Retired Administrative Law
Judge, Federal Trade Commission, Wheaton, Md.
HERBERT GROSSMAN, ASLBP Attorney, Bethesda,
Md.*
DR. DAVID B. HALL, Los Alamos Scientific Laboratory,Los, Alamos, N.M.
DR. CADET H. HAND, JR., Director, Bodega Marine
Laboratory, University of California, Bodega Bay, Calif.
DR. DAVID L. HETRICK, Professor, Nuclear Engineering
Department, University of Arizona, Tucson, Ariz.
ERNEST E. HILL, Engineer, Lawrence Livermore
Laboratory, University of California, Livermore, Calif.
DR. ROBERT L. HOLTON, School of Oceanography,
Oregon State University, Corvallis, Ore.
DR. FRANK F. HOOPER, Chairman, Resource Ecology
Program, School of Natural Resources, University ofMichigan, Ann Arbor, Mich.
ELIZABETH B. JOHNSON, Engineer, Oak Ridge National
Laboratory, Oak Ridge, Tenn.
DR. WALTER H. JORDAN, Retired Senior Research Advisor & Physicist, Oak Ridge National Laboratory, Oak
Ridge, Tenn.
DR. JAMES C. LAMB, III, Department of Environmental
Sciences & Engineering, University of North Carolina,
Chapel Hill, N.C.
MARGARET M. LAURENCE, Partner, law firm ofLaurence, Stokes and Neilan, Arlington, Va.
DR. J. V. LEEDS, JR., Professor, Environmental and Electrical Engineering, Rice University, Houston, Tex.
GUSTAVE A. LINENBERGER, ASLBP Physicist, Bethesda"
Md.*
DR. LINDA W. LITTLE, Research Triangle Institute,
Research Triangle Park, N.C. Department of Environmental Sciences & Engineering, University of North Carolina,
Chapel Hill, N.C.
DR. M. STANLEY LIVINGSTON, Retired Associate Director, Atomic Energy Commission National Accelerator
Laboratory, Santa Fe, N.M.
DR. EMMETH A. LUEBKE, ASLBP Physicist, Bethesda,
Md.*
DR. WILLIAM E. MARTIN, Senior Ecologist, Battelle
Memorial Institute, Columbus, Ohio
DR. KENNETH A. McCOLLOM, Dean, Division ofEngineering, Technology and Architecture, Oklahoma
State University, Stillwater, Okla.
GARY L. MILHOLLIN, University of Wisconsin Law
School, Madison, Wis.
MARSHALL E. MILLER, ASLBP Attorney, Bethesda,
Md.'"
DR. OSCAR H. PARIS, ASLBP Environmental Scientist,
Bethesda, Md. '"DR. HUGH PAXTON, Los Alamos Scientific Laboratory,
Los Alamos, N.M.

DR. PAUL W. PURDOM, Director, Environmental Studies
Institute, Drexel University, Philadelphia, Pa.
DR. FORREST J. REMICK, Director, Institute of Scienceand Engineering, Pennsylvania State University, University Park, Pa.
DR. DAVID R. SCHINK, Department of Oceanography,
Texas A&M University, College Station, Tex.
FREDERICK J. SHON, ASLBP Physicist, Bethesda, Md. *IVAN W. SMITH, Administrative Law Judge, U.S. Nuclear
Regulatory Commission, Bethesda, Md. '"DR. MARTIN J. STEINDLER, Chemist, Argonne National
Laboratory, Argonne, Ill.
DR. QUENTIN J. STOBER, Research Associate Professor,
Fisheries Research Institute, University of Washington,
Seattle, Wash.
JOSEPH F. TUBRIDY, Attorney at Law, Washington, D.C.
SEYMOUR WENNER, Retired Administrative Law Judge,
Postal Rate Commission, Washington, D.C.
JOHN F. WOLF, Attorney, law firm of Lamensdorf,
Leonard & Moore, Washington, D.C.
SHELDON J. WOLFE, ASLBP Attorney, Bethesda, Md. *Atomic Safety and Licensing Appeal Panel
An Atomic Safety and Licensing Appeal Board, establishedeffective September 18, 1969, was delegated the authority toperform the review function which would otherwise be performed by the Commission in proceedings on applications forlicenses or authorizations in which the Commission had adirect financial interest, and in such· other licensing proceedings as the Commission might specify.
In view of the increase in the number of proceedings subject to administrative appellate review, the Atomic Safety
and Licensing Appeal Panel was established on October 25,
1972, from whose membership three-member appeal boards
could be designated for each proceeding in which the Commission had delegated its authority to an appeal board. At thesame time, the Commission modified its rules to delegate
authority to appeal boards in all proceedings involving thelicensing of production and utilization facilities (for example,
power reactors).
Pusuant to subsection 201 (g)(l) of the Energy Reorganization Act of 1974, the functions performed by appeal boards
were specifically transferred to the Nuclear Regulatory Commission. The Commission appoints members to the Appeal
Panel, and the Chairman of the panel (or, in his absence, theVice Chairman) designates a three-member appeal board foreach proceeding. The Commission retains review authority
over decisions and actions of appeal boards. The appeal
board panel, on September 30, 1979, was composed of thefollowing full-time members and professional staff:
ALAN S. ROSENTHAL, Appeal Panel Chairman, U.S.
Nuclear Regulatory Commission, Bethesda, Md.
DR. JOHN H. BUCK, Appeal Panel Vice Chairman, U.S.
Nuclear Regulatory Commission, Bethesda, Md.
MICHAEL C. FARRAR, Appeal Panel Member, U.S.
Nuclear Regulatory Commission, Bethesda, Md.
RICHARD S. SALZMAN, Appeal Panel Member, U.S.
Nuclear Regulatory Commission, Bethesda, Md.
JOHN CHO, Counsel, Appeal Panel, U.S. Nuclear
Regulatory Commission, Bethesda, Md.

288

CARDIS L. ALLEN, Technical Advisor, Appeal Panel, U.S.
Nuclear Regulatory Commission, Bethesda, Md.
LINDA S. GILBERT, Legal Intern, Appeal Panel, U.S.
Nuclear Regulatory Commission, Bethesda, Md.
ROBERT S. PERLIS, Legal Intern, Appeal Panel, U.S.
Nuclear Regulatory Commission, Bethesda, Md.
During fiscal year 1979, the Appeal Panel also included thefollowing part-time members:
DR. LAWRENCE R. QUARLES, Dean Emeritus, School ofEngineering and Applied Science, University of Virginia,
Charlottesville, Va.
DR. W. REED JOHNSON', Professor of Nuclear Engineering, University of Virginia, Charlottesville, Va.
Advisory Committee on Medical Uses of Isotopes
The Advisory Committee on Medical Uses of Isotopes wasestablished in July 1958. The ACMI, composed of qualified
physicians and scientists, considers medical questions referred
to it by the NRC staff, and renders expert opinion regarding
medical use of radioisotopes. The ACMI also advises the NRC
staff, as requested, on matters of policy. Members are
employed under yearly personal services contracts. The
Deputy Director, Division of Fuel Cycle and Material Safety,
serves as Committee Chairman. As of September 30, 1979 themembers were:

RICHARD E. CUNNINGHAM, Chairman, ACMI, Deputy
Director, Division of Fuel Cycle and Material Safety, U.S.
Nuclear Regulatory Commission, Silver Spring, Md.
DR. FRANK H. DE LAND, Chief" Nuclear Medicine
Department, Veterans' Administration Hospital, Lexington, Ky.
DR. EDWARD W. WEBSTER, Director, Department ofRadiation Physics, Massachusetts General Hospital,
Boston, Mass.
DR. JOSEPH B. WORKMAN, Associate Professor ofRadiology, Duke University Medical Center, Durham,
N.C.
DR. VINCENT P. COLLINS, Medical Director, Houston
Institute for Cancer Research, Diagnosis and Treatment,
Houston, Tex.
DR. MELVIN L. GRIEM, Professor and Director, Chicago
Tumor Institute, University of Chicago, Chicago, Ill.
DR. SALLY DENARDO, Director, Nuclear HematologyOncology, Department of Nuclear Medicine, University ofCalifornia-Davis Medical Center, Sacramento, Ca.
DR. JACK GOODRICH, Radiology Associates of Erie,
Hamot Medical Center, Erie, Pa.
DR. B. LEONARD HOLMAN, Chief, Clinical Nuclear
Medicine, Department of Radiology, Peter Bent Brigham
Hospital, Boston, Ma.
DR. DAVID H. WOODBURY, Director, Nuclear Medicine,
Wayne County General Hospital, Eloise, Mi.

=================================================289
Appendix 3

Public Document Rooms
Most documents originated by NRC, or submitted to it for consideration, are placed in the Commission's Public Document Roomat 1717 H Street, N.W., Washington, D.C., for public inspection. In addition, documents relating to licensing proceedings or
licensed operation of specific facilities are made available in local public document rooms established in the vicinity of each proposed or existing nuclear facility. The locations of these local PDRs as of December 1979, and the name of the facility for whichdocuments are retained, are listed below. (NOTE: Updated listings of local PDRs may be obtained by writing to the Local PublicDocument Room Branch, Division of Rules and Records, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555.)

ALABAMA• Mrs. Maude S. Miller
Athens Pubic Library
South and Forrest
Athens, Ala. 35611
Browns Ferry Nuclear Plant• Mr. Wayne Love
G.S. Houston Memorial Library
212 W. Burdeshaw Street
Dothan, Ala. 36303
Farley Nuclear Plant• Mrs. Joanne Wyatt
Clanton Public Library
100 First Street
Clanton, Ala. 35045
Barton Nuclear Plant• Mrs. Peggy McCutchen
Scottsboro Public Library
1002 South Broad Street
Scottsboro, Ala. 35768
Bellefonte Nuclear Plant

ARIZONA• Mrs. Mary Carlson
Phoenix Public Library
Science and Industry Section
12 East McDowell Road
Phoenix, Arix. 85004
Palo Verde Nuclear Plant

ARKANSAS• Mr. Vaughn
Arkansas Polytechnic College
Russellville, Ark. 72801
Arkansas Nuclear One• Mrs. Alice Rosenberger
Palo Verde Valley District Library
125 West Chanslorway
Blythe, Calif. 92255• Mr. William B. Rohan
San Diego County Law Library
1105 Front Street
San Diego, Calif. 92101
Sundesert Nuclear Plant• Mrs. Eileen Danforth
Mission Viejo Branch Library
24851 Chrisanta Drive
Mission Viejo, Calif. 92676
San Onofre Nuclear Plant• Mr. Chi Su Kim
Documents and Maps Department
California Polytechnic State
University Library
San Luis Obispo, Calif. 93407
Diablo Canyon Nuclear Plant• Mrs. Judy Klapprott
Humboldt County Library
636 F Street
Eureka, Calif. 95501
Humboldt Bay Nuclear Plant• Mrs. Dorothy Harvey
Business & Municipal Department
Sacramento City-County Library
828 I Street
Sacramento, Calif. 95814
Rancho Seco Nuclear Plant• Stanislaus County Free Library
1500 I Street
Modesto, Calif. 95345
Stanislaus Nuclear Plant

CALIFORNIACOLORADO• Mr. C. Combs
Kern County Library
1315 Truxtun Avenue
Bakersfield, Calif. 93301
San Joaquin Nuclear Plant• Miss Ester Fromm
Greeley Public Library
City Complex Building
Greeley, Colo. 80631
Fort St. Vrain Nuclear Plant• Mrs. Elizabeth Morrissett
Acquisitions Department
Auraria Library
University of Colorado at Denver
Lawrence and 11th
Denver, Colo. 80204
Atlas Corp. Uranium MillCONNECTICUT• Mrs. Judy Liskov
Waterford Public Library
Rope Ferry Road-Route 156
Waterford, Conn. 06385
Millstone Nuclear Plant• Mr. Stewart Porter
Russell Library
119 Broad Street
Middletown, Conn. 06457
Haddam Neck Nuclear Plant

DELAWARE• Mrs. Yvonne Puffer
Newark Free Library
750 East Delaware Avenue
Newark, Del. 19711
Summit Nuclear Plant

FLORIDA• Ms. Sally Litton
Jacksonville Public Library
122 North Ocean Street
Jacksonville, Fla. 32204
Offshore Power Systems
Manufacturing Facility• Mrs. R. Scott
Indian River Community College
Library3209 Virginia Avenue
Ft. Pierce, Fla. 33450
St. Lucie Nuclear Plant• Mrs. Rene' Daily
Environmental and Urban
Affairs Library

290

Florida International University
Miami, Fla. 33199
Turkey Point Nuclear Plant• Mrs. Bonsall
Crystal River Public Library
668 N.W. First
Crystal River, Fla. 32639
Crystal River Nuclear Plant

GEORGIA• Mrs. J. W. Borom
Burke County Library
Fourth Street
Waynesboro, Ga. 30830
Vogtle Nuclear Plant• Ms. Annette Osborne
Appling County Public Library
103 City Hall Drive
Baxley, Ga. 31513
Hatch Nuclear Plant

ILLINOIS• Mr. Ed Anderson
Illinois Valley Community College
Rural Route #1
Oglesby, Ill. 16348
LaSalle Nuclear Plant• Mrs. Pam Wilson
Morris Public Library
604 Liberty Street
Morris, Ill. 60451
Dresden Nuclear Plant
Midwest Fuel Recovery Plant• Mrs. Marie Hoschied
Moline Public Library
504 17th Street
Moline, Ill. 61255
Quad Cities Nuclear Plant• Ms. Jo Ann Ellingson
Zion~Benton Publici Library
2600 Emmaus Avenue
Zion, Ill. 60099
Zion Nuclear Plant• Mrs. M. Evans
Vespasian Warner Public Library
120 West Johnson Street
Clinton, Ill. 61727
Clinton Nuclear Plant• Mrs. Penny O'Roarke
Byron Public Library
Third and Washington Streets
Byron, Ill. 61010
Byron Nuclear Plant• Mr. Thomas Carter
Wilmington Township Public Library
201 S. Kankakee Street
Wilmington, Ill. 60481
Braidwood Nuclear Plant• Savanna Township Public Library
326 Third Street
Savanna, Ill. 61074
Carroll Nuclear Plant

INDIANA.' West Chester Township Public
Library
125 South Second Street
Chestertown, Ind. 46304
Bailly Nuclear Plant• Ms. Carol Cowles
Madison-Jefferson County Public
Library
420 West Main Street
Mad~on, Ind. 47250
Marble Hill Nuclear Plant

IOWA• Miss Kay Burke
Reference Service
Cedar Rapids Public Library
428 Third Avenue, S.E.
Cedar Rapids, la. 52401
Duane Arnold Nuclear Plant

KANSAS• Mr. Jack Scott
Coffey County Courthouse
Burlington, Kans. 66839
Wolf Creek Nuclear Plant

KENTUCKY• Mr. Clarence R. Graham
Louisville Free Public Library
4th and York Streets
Louisville, Ky. 40203
Marble Hill Nuclear Plant

High Street
W~~et, Me. 04578

Maine Yankee Nuclear Plant

MARYLAND• Mrs. Elizabeth Hart
Charles County Library
Garrett and Charles Streets
La Plata, Md. 20646
Douglas Point Nuclear Plant• Mrs. Marie Barrett
Calvert County Library
Prince Frederick, Md. 20678
Calvert Cliffs Nuclear Plant• Ms. Pamela R. Schott
Harford Community College
401 Thomas Run Road
Bel Air, Md. 21014
Perryman Nuclear Plant

MASSACHUSETTS• Mrs. Margaret Howland
Greenfield Community College
One College Drive
Greenfield, M~. 01301
Yankee Rowe Nuclear Plant• Mr. Mark Titus
Plymouth Public Library
North Street
Plymouth, M~. 02360
Pilgrim Nuclear Plant• The Carnegie Library
Avenue A
Turner Falls, M~. 01376
Montague Nuclear Plant

LOUISIANA• Business & Science Division
New Orleans Public Library
219 Loyola Avenue
New Orleans, La. 70140
Offshore Power Systems
Manufacturing Facility• Mr. Ken Owen
University of New Orleans LibraryLouisiana Collection, Lakefront
New Orleans, La. 70122
Waterford Nuclear Plant• Mrs. Freeda Fisher
Audubon LibraryWest Feliciana Branch
Ferdinand Street ,St. Francisville, La. 70775• Mr. Jimmie H. Hoover
Government Documents Department
Louisiana State University
Baton Rouge, La. 70803
River Bend Nuclear Plant

MAINE• Mrs. Barbara Shelton
Wiscasset Public Library

MICHIGAN• Mrs. Diana Shamp
Reference Department
Kalamazoo Public Library
315 South Rose Street
Kalamazoo, Mich. 49006
Palisades Nuclear Plant• Mrs. Katherine Thomson
St. Clair County Library
210 McMorran BoulevardPort Huron, Mich. 48060
Greenwood Nuclear Plant• Mrs. M. B. Wallick
Charlevoix Public Library
107 Clinton Street
Charlevoix, Mich. 49720
Big Rock Point• Mrs. Averill Packard
Grace Dow Memorial Library1710 West St. Andrews Road
Midland, Mich. 48640
Midland Nuclear Plant• Ms. Ann Stobbe
Maude Preston Palenske
Memorial Library291

500 Market Street
St. Joseph, Mich. 49085
D.C. Cook Nuclear Plant• Mrs. Marcia Learned
Reference Department
Monroe County Library System
3700 South Custer Road
Monroe, Mich. 48161
Fermi Nuclear Plant

MINNESOTA• Mrs. Copeland
Environmental Conservation Library
Minneapolis Public Library
300 Nicollet Mall
Minneapolis, Minn. 55401
Monticello Nuclear Plant
Prairie Island Nuclear Plant

MISSOURI• Mrs. Ladonna Justice
Fulton City Library
709 Market Street
Fulton, Mo. 65251• Mrs. Ranata Rotkowicz
Olin Library of Washington
University
Skinker & Lindell Boulevard
St. Louis, Mo. 63130
Callaway Nuclear Plant

MISSISSIPPI• Mrs. Stella Jennings
Clairborne County Chancery Clerk
Clairborne County Courthouse
Port Gibson, Miss. 39150
Grand Gulf Nuclear Plant• Mr. William McMullin
Corinth Public Library
1023 Fillmore Street
Corinth, Miss. 38834
Yellow Creek Nuclear Plant

NEBRASKA• Mr. Frank Gibson
W. Dale Clark Library
215 South 15th Street
Omaha, Neb. 68102
Ft. Calhoun Nuclear Plant• Mrs. Loy Mowery
Auburn Public Library
118 15th Street
Auburn, Neb. 68305
Cooper Nuclear Plant

NEW HAMPSHIRE• Miss Pamela Gjettum
Exeter Public Library
Front Street
Exeter, N. H. 03883
Seabrook Nuclear Plant

NEW JERSEY• Stockton State College Library
Pomona, N.J. 08240
Offshore Power Systems
Manufacturing Facility
Atlantic Nuclear Plant• Miss Elizabeth Fogg
Salem Free Public Library
112 West Broadway
Salem, N.J. 08097
Salem Nuclear Plant
Hope Creek Nuclear Plant• Mrs. Gail Colure
Ocean County Library
Brick Township Branch
401 Chambers Bridge Road
Brick Town, N.J. 08723
Oyster Creek Nuclear Plant
Forked River Nuclear Plant

NEW MEXICO• Ms. Sandra Coleman
Ge~eral Library, Reference
Department
University of New Mexico
Albuquerque. N.M. 87131• Ms. Ingrid Vollnhofer
New Mexico State LibraryBox 1629
Santa Fe, N.M. 87503
Waste Isolation Pilot Plant• Mrs. Dorothy Augustine
Catskill Public Library
One Franklin Street
Catskill, N.Y. 12414
Greene County Nuclear Plant• Mr. Stanley Zukowzki
Buffalo & Erie County Public
Library
Lafayette Square
Buffalo, N.Y. 14203• Ms. Marsha Russell
Town of Concord Public Library
23 North Buffalo Street
Springville, N.Y. 14141
NFS Fuel Reprocessing Plant andUFa Facility• Mr. Sol Becker
Public Health Library
New York City
Department of Health
125 Worth Street
New York, N.Y. 10013
Columbia University
Research Reactor• Mr. Harold Ettelt
Columbia-Greene Community
College
P.O. Box 100
Hudson, N.Y. 12534
Greene County Nuclear Plant

NORTH CAROLINANEW YORK• Mr. Ralph W. Schmidt
Oswego County Office Building
46 East Bridge Street
Oswego, N.Y. 13126
Nine Mile Point Nuclear Plant
Sterling Nuclear Plant
Fitzpatrick Nuclear Plant• Mrs. June Rogoff
Rochester Public LibraryBusiness & Social Science Division
115 South Avenue
Rochester, N.Y. 14604
Ginna Nuclear Plant• Mr. Oliver SwiftWhite Plains Public Library
100 Martine Avenue
White Plains, N.Y. 10601
Indian Point Nuclear Plant• Shoreham-Wading River Public
Library
Route 25A
Shoreham, N.Y. 11786
Shoreham Nuclear Plant• Mrs. E. Overton
Riverhead Free Library
330 Court Street
Riverhead, N.Y. 11901
Jamesport Nuclear Plant• Mrs. Ruth Osborne
Public Library of Charlotte &Mecklenburg County
310 North Tryon Street
Charlotte, N.C. 28202
McGuire Nuclear Plant• Mr. Roy Dicks
Wake County Public Library
104 Feyetteville Street
Raliegh, N.C. 27601
Shearon Harris Nuclear Plant• Mr. David G. Ferguson
Davie County Public Library
416 North Main Street
P.O. Box 158
Mocksville, N.C. 27028
Perkins Nuclear Plant• Southport-Brunswick County Library
109 West Moore Street
Southport, N.C. 28461
Brunswick Nuclear Plant• Mrs. Charlotte Ellis
Franklin County Library
1026 Justice Street
Louisburg, N.C. 27549
Gulf Youngsville Fuel Fabrication
Facility

292

OHIO
• Mrs. Betty Waltman
Perry Public Library
3753 Main Street
Perry, Ohio 44081
Perry Nuclear Plant• Clermont County Library
Third and Broadway Streets
Batavia, Ohio 45103
Zimmer Nuclear Plant• Mr. Donald Fought
Ida Rupp Public Library
310 Madison Street
Port Clinton, Ohio 43452
Davis-Besse Nuclear Plant• Mrs. Esther Schedley'
Berlin Township Public Library
Four East Main Street
Berlin Heights, Ohio 44814
Erie Nuclear Plant

OKLAHOMA• Mr. Craig Buthod
Tulsa City-County Library
400 Civic Centeri Tulsa, Okla. 74102
Black Fox Nuclear Plant• Mrs. O.J. Grosclaude
Sallisaw City Library
III North Elm
Sallisaw, Okla. 74955
Sequoyah UFo Facility• Ms. Hazel Nicholson
Guthrie Public Library
402 East Oklahoma Street
Guthrie, Okla. 73044
Cimarron Pu Fabrication Plant
and Uranium Fuel Facility

OREGON• Mr. H. B. Allen
City Hall, Records Office
Arlington, Ore. 97812
Pebble Springs Nuclear Plant• Mr. Zimmer
Columbia County Courthouse
Law Library Circuit Court Room
St. Helens, Ore. 97501
Trojan Nuclear Plant

PENNSYL VANIA• Reference Department
Osterhout Free Library
71 South Franklin Street
Wilkes-Barre, Pa. 18701
Susquehanna Nuclear Plant• Mrs. Margaret Atwood
York College of Pennsylvania
Country Club Road
York, Pa. 17405
Three Mile Island Nuclear Plant• Mr. John Geschwindt
Government Publications Section
State Library of Pennsylvania
Education Building
Commonwealth and Walnut Street
Harrisburg, Pa. 17126
Peach Bottom Nuclear Plant
Three Mile Island Nuclear Plant
Fulton Nuclear Plant• Mrs.' Gordon Bauerle
Pottstown Public Library
500 High Street
Pottstown, Pa. 19464
Limerick Nuclear Plant• Apollo Memorial Library
219 North Pennsylvania Avenue
Apollo, Pa. 15613
Apollo UFo and Pu Facilities• Mr. Anthony Martin
Carnegie Library of Pittsburgh
4400 Forbes Avenue
Pittsburgh, Pa. 15213
Cheswick Fuel Development
Laboratories• Mr. F.E. Virostek
B.F. Jones Memorial Library
663 Franklin Avenue
Aliquippa, Pa. 15001
Beaver Valley Nuclear Plant .
Shippingport Light Water Breeder
Reactor

PUERTO RICO
• Mrs. Rosario Cabrera
Public Library, City HallJose de Diego Avenue
P.O. Box 1086
Arecibo, P.R. 00612• Mrs. Amalia Ruiz De Porras
Etien Totti Public Library
College of Engineers, Architects& Surveyors
Urb Roosevelt Development
Hato Rey, P.R. 00918
North Coast Nuclear Plant

RHODE ISLAND• Mrs. Ann Crawford
Cross Mill Public Library
Old Post Road
Charlestown, R.I. 02831• Mr. Thomas Reynolds
University of Rhode Island
University Library
Government Publications Office
Kingston, R.I. 02881
New England Nuclear Plant

SOUTH CAROLINA• Joe E. Garcia
York County Library

325 South Oakland Avenue
Rock Hill, S.C. 29730
Catawba Nuclear Plant• Reference Department
Richland County Public Library
1400 Sumter Street
Columbia, S.C. 29201
Summer Nuclear Plant• Miss Louise Marcum
Oconee County Library
501 W. Southbroad
Walhalla, S.C. 29691
Oconee Nuclear Plant• Mrs. Allene Reep
Hartsville Memorial LibraryHome and Fifth Avenues
Hartsville, S.C. 29550
H.B. Robinson Nuclear Plant• Mr. David Eden
Cherokee County Library
300 East Rutledge Avenue
Gaffney, S.C. 29340
Cherokee Nuclear Plant• Mr. Fred Bodiford
County Office Builidng
Room 105
P.O. Box 443
Barnwell, S. C. 29812
Barnwell Fuel Plant
UFo Facility
Barnwell Fuel Storage Station• Mr. Carl Stone
Anderson County Library
202 East Greenville Street
Anderson, S.C. 29621
Recycle Fuel Plant• Mrs. Ellen Jenkins
Barnwell County Library
Hagood Avenue
Barnwell, S.C. 29812
Chern-Nuclear Plant

TENNESSEE• Miss Kendall J. Cram
Tennessee State Library and Archives
403 Seventh Avenue, North
Nashville, Tenn. 37219
Hartsville Nuclear Plant• Ms. Dorothy Dismuke
Oak Ridge Public Library
Civic Center
Oak Ridge, Tenn. 37830• Mrs. Patricia Rugg
Lawson McGhee Public Library
500 West Church Street
Knoxville, Tenn. 37902
Clinch River Breeder Plant
Exxon Nuclear Fuel Recovery
CenterFuel Fabrication Facility• Mr. Wally Keasler
Chattanooga-Hamilton County
Bicentennial Library293

1001 Broad Street
Chattanooga, Tenn. 37402
Sequoyah Nuclear Plant
Watts Bar Nuclear Plant• Mr. T. Cal Hendrix
Kingsport Public Library
Broad and New Streets
Kingsport, Tenn. 37660
Phipps Bend Nuclear Plant• Mr. H.E. Zittel
Oak Ridge National Laboratory
P.O. Box X
Oak Ridge, Tenn. 37830
Tyrone Nuclear Plant

TEXAS• Mrs. Tim Whitworth
Somervell County Public Library
On The Square
P.O. Box 1417
Glen Rose, Tex. 76043
Comanche Peak Nuclear Plant• Newton County Library
P.O. Box 657
Newton, Tex. 77034
Blue Hills Nuclear Plant• Matagorda County Courthouse
Matagorda County Law Library
P.O. Box 487
Bay City, Tex. 77414
South Texas Nuclear Plant• Mrs. Kroesche
Sealy Public Library
201 Atchison Street
Sealy, Tex. 77474
AlIens Creek Nuclear Plant

VIRGINIA• Ms. Sandra Peterson
Swem Library
College of William & Mary
Williamsburg, Vlit. 23185
Surry Nuclear Plant• Ms. Manrique
Board of Supervisors
Louisa County Courthouse
P.O. Box 27
Louisa, Va. 23093• Mr. Gregory Johnson
Alderman Library
Manuscripts Department
University of Virginia
Charlottesville, Va.' 22901
North Anna Nuclear Plant

WASHINGTON• Ms. D. E. Roberts
Richland Public Library
Swift and Northgate Streets
Richland, Wash. 99352
WPPSS 1, 2 and 4 Nuclear Plants
Exxon Fuel Plant• Mrs. D. Stendal
Sedro Wooley Library
802 Ball Avenue
Sedro Wooley, Wash. 98294
Skagit Nuclear Plant• Ms. Selma Nielsen
W. H. Abel Memorial Library
125 Main Street South
Montesano, Wash. 98563
WPPSS 3 and 5 Nuclear Plants

VERMONTWISCONSIN• Mrs. June Bryant
Brooks Memorial Library
224 Main Street
Brattleboro, Vt. 05301
Vermont Yankee Nuclear Plant• Mrs. Jane Radloff
LaCrosse Public Library
800 Main Street
LaCrosse, Wis. 54601
LaCrosse BWR Nuclear Plant• Mr. Arthur M. Fish
Document Department, Library
University of Wisconsin
Stevens Point
Stevens Point, Wis. 54481
Point Beach Nuclear Plant
Wood Nuclear Plant• Mead Public Library
710 North Eighth Street
Sheboygan, Wis. 53081• Madison Public LibraryBusiness and Science Division
201 West Mifflen Street
Madison, Wis. 53703
Haven Nuclear Plant• Ms. Sue Grossheuch
Kewaunee Public Library
822 Juneau Street
Kewaunee, Wis. 54216
Kewaunee Nuclear Plant• Mr. John Jax
University of Wisconsin
Stout Library
Menomonie, Wis. 54751• Mr. Robert Fetvedt
University Library
University of Wisconsin-Eau Claire
Park and Garfield Avenues
Eau Claire, Wis. 54710• Mrs. Robert Goodrich
Durand Free Library
315 Second Avenue, WestDurand, Wis. 54736
Tyrone Nuclear Plant

WYOMING• Mrs. Carroll Highfill
Converse County Library
Douglas, Wyo. 82633
Highland Uranium Mill• Mrs. Margaret Baker
Carbon County Public Library
Courthouse
Rawlins, Wyo. 82301
Shirley Basin Uranium Mill

294===================================================
Appendix 4

Regulations and Amendments-Fiscal Year 1979

The regulations of the Nuclear Regulatory Commission are contained in Title 10, Chapter 1, of the Code of Federal Regulations. Effective and proposed regulations concerning licensed activities, and certain policy statements relating thereto, whichwere published in the Federal Register during fiscal year 1979, are described briefly below.

REGULATIONS AND AMENDMENTS PUT INTO EFFECT
Distribution of Applications and Environmental Statements
to Local Officials-Parts 2 and 51
On October 6, 1978, amendments to Parts 2 and 51 were
published, effective November 6, 1978, to provide for noticeto the chief executives of the appropriate alternative
municipalities or counties which have been identified in theapplication or environmental report as alternative sites fornuclear facilities or activities.

Radiation Surveys of Therapy Patients-Part 35
On November 28, 1978, an amendment to Part 35 waspublished, effective December 28, 1978, to require licensees
authorized to treat patients with temporary implants incorporating radioactive material to confirm the removal of theimplants at the end of the treatment by (1) a source count and(2) a radiation survey of the patient.
Codes and Standards for Nuclear Power Plants-Part 50

Amendments Regarding Basic Component-Part 21
On October 19, 1978, amendments to Part 21 were published, effective immediately, to limit the types of items that arewithin the scope of the NRC rule for reporting defects andnoncompliance.

On November 30, 1978, amendments to Part 50 were
published, effective immediately, to clarify certain ambiguities to avoid misinterpretations of provisions which deal
with requirements for in service inspection of nuclear power
plants.
Calibration of Teletherapy Units-Part 35

Revocation of Certain Reporting Requirements-Part 50
On October 25, 1978, amendments to Part 50 were published, effective immediately, to revoke two reporting requirements on antitrust information and to revoke a requirement for retention of 25 copies of the antitrust report during
the antitrust review.
Standards for Combustible Gas Control Systems-Part 50
On October 27, 1978, amendments to Part 50 were published, effective November 27, 1978, to clarify the NRC's position on its general design criterion regarding the containment
design basis and to provide a new section specifying the standards for combustible gas control systems.
Miscellaneous Amendments-Parts 20, 21, 40, and 73
On November 9, 1978, amendments to Parts 20, 21,40, and73, were published, effective immediately, to change thetelephone number for the NRC's Inspection and Enforcement
Regional Office I and to exempt the general licensee under§40.25 from the requirements of the NRC's regulation"Notices, Instructions, and Reports to Workers; Inspection."

On January 8, 1979, amendments to Part 35 were published,effective July 9, 1979, to require medical licensees to (1)calibrate each teletherapy unit annually and (2) perform
monthly spot checks on those calibrations.
Miscellaneous Amendments-Parts 20, 21, and 73
On January 12, 1979, amendments to Parts 20, 21, and 73
were published, effective immediately, which change thetelephone number for the Commission's Inspection and Enforcement Regional Office V.
Clarification of Computation of Time Provisions-Part 2
On January 22, 1979, amendments to Part 2 were published,effective immediately, to clarify the requirements for computing the time prescribed for replying to a petition for leaveto intervene.
Change in License Conditions for Certain Medical
Licenses-Part 35
On February 20, 1979, an amendment to Part 35 waspublished, effective March 22, 1979, (a) to permit physicians
greater latitude, when they use certain low risk diagnostic

===================================================295
radiopharmaceuticals, by no longer designating authorized
clinical procedures and (b) by deleting from several licensing
groups certain chemical forms not approved by FDA.
Requirements for the Physical Protection of Nuclear Power
Plants-Part 73
On February 28, 1979, an amendment to Part 73 waspublished, effective immediately, to change from February
23, 1979 to August 1, 1979, the implementation date when
pat-down searches of regular employees at nuclear power
plants, the two man rule procedures, and compartmentalization have to be implemented for protection against insider
sabotage.
Waiver or Reduction of Fees for Searching and Reproductionof Records-Part 9

Physical Protection of Irradiated Reactor Fuel in Transit-Part 73
On June 15, 1979, amendments to Part 73 were published, effective July 16, 1979, which would establish requirements forprotection of spent fuel in transit.
Conduct of Employees Ownership of Stocks, Bonds, andOther Security Interests by NRC Employees-Part 0
On July 17, 1979, amendments to Part 0 were published, effective immediately, in which the NRC revised its regulations
governing the ownership of stocks, bonds, and other security
interests.
Safeguards Requirements for Special Nuclear Material ofModerate and Low Strategic Significance-Parts 70,73, and150

On March 16, 1979, an amendment to Part 9 was published,eff~tive April 16, 1979 to add a new section "Waiver or
Reduction of Fees."

On July 24, 1979, amendments to Parts 70, 73, and 150 were
published, effective November 21, 1979, to require physicalprotection measures to detect theft of special nuclear material.of moderate and low strategic significance.

Rules of General Applicability to Domestic Licensing of
Byproduct Material, Domestic Licensing of Source Material
and Domestic Licensing of Special Nuclear Material-Parts
30,40 and 70

Rules of Practice; Selected Nuclear Power Plant Construction
Permit Applications-Part 2

On March 22, 1979, an amendment to Parts 30, 40, and 70
was published, effective June 5, 1979, to require licensees to
notify the Commission when they deCide to permanently
discontinue all activities involving materials authorizedunder a license. This will allow NRC to terminate the license
in an orderly and timely manner.

On August 15, 1979, amendments to Part 2 were published,effective immediately, to permit the use of new staff procedures on a trial basis for the systems and site safety portionsof selected nuclear power plant construction permit applications.
Requirements for the Physical Protection of Nuclear Power
Plants-Part 73

Financial Protection Requirements and Indemnity
Agreements; Miscellaneous Amendments-Part 140

On August 15, 1979, amendment to Part 73 was published,effective immediately, to change the date of the pat-down
searches of regular employees at nuclear power plants from
August 1, 1979 to November 1, 1979.

On April 6, 1979, amendments to Part 140 were published,effective May 1, 1979, to increase the level of the primarylayer of financial protection required of certain indemnified
licensees.

Licensing of Production and Utilization Facilities; and Accessfor Resident Inspection-Parts 50 and 70

General License Requirements For Any Person Who Possesses
Formula Quantitites of Strategic Special Nuclear Material(SSNM) In Transit Subject to Certain Requirements-Part
170
On May 8, 1979, amendments to Part 70 were published, effective June 7, 1979, to remove the exemptions to licensing for
carriers and other persons who possess or control formulaquantities of strategic special nuclear material for the purposeof transport, or storage incident of transport.
Control of Radiation Exposure to Transient Workers-Parts
19 and 20
On June 6, 1979, amendments to Parts 19 and 20 were
published, effective August 20, 1979, which would require
NRC licensees to control the total occupational radiation doseof individuals who work in NRC~1icensed activities.

On August 16, 1979, amendments to Parts 50 and 70 were
published, effective September 17, 1979, to require powerreactor licensees, construction permit holders, and selectedfuel facility licensees to provide (1) on site, rent-free, exclusive use of office space and (2) immediate licensee facility
access to Commission inspection personnel.
Uranium Mill Tailings Licensing-Parts 40 and 150
On August 24, 1979, amendments to Part 40 and 150 were
published, effective immediately, to conform to the requirements of the Uranium Mill Tailings Radiation Control
Act of 1978 and to the standards set forth in the draft Generic
Environmental Impact Statement on Uranium Milling.
Addition of Veterinarians to the In Vitro General
License-Parts 31 and 32
On August 28, 1979, amendments to Parts 31 and 32 were
published, effective September 27, 1979, to add veterinarians

296

to the groups already authorized to use byproduct material
under general license for clinical or laboratory testing done
outside the body.
Licenses for Radiography and Radiography Safety Requirements for Radiographic Operations; Amendments ofRadiography Regulations-Part 34
On August 30, 1979, amendments to Part 34 were published,effective March 3, 1980, to require several changes intended
to improve radiography safety and to formalize as regulations
current licensing practices.
Amendments to Appendix A-Requests for Declassification
Review-Part 9
On August 30, 1979 amendments to Part 9 were published effective immediately, to substitute references to E.O. 12065
and its implementing directive (Information Security Oversight office Directive No.1) for the replaced E.O. 11652 andits implementing National Security Directive.

REGULATIONS AND AMENDMENTS PROPOSED
Storage of Spent Fuel in an Independent Spent Fuel Storage
Installation-Part 72
On October 6, 1978, proposed amendments to Part 72 were
published for comment to add a new part which specifies procedures and requirements for issuance of licenses to store
spent fuel in an independent spent fuel storage installation.
Management and Disposal of Low-Level Wastes by Shallow
Land Burial and Alternative Disposal Methods-Part 61
On October 25, 1978, an advance notice of proposed
rulemaking was published for comment to add a new Part 61
which specifies a regulatory program for management of
low-level radioactive wastes.
Licensing Procedures for Geologic Repositories for HighLevel Radioactive Wastes-Proposed General Statement of
PolicyOn November 17, 1978, a proposed general statement ofpolicy was published for comment regarding establishment ofprocedures for licensing geologic high-level waste repositories
to be constructed and operated by the U.S. Department ofEnergy.
Burial of Small Quantities of Radionuclides-Part 20
On December 4, 1978, proposed amendments to Part 20 were
published for comment to require NRC licensees to obtain
Commission approval prior to burial of small quantities of radionuclides.
Acceptance Criteria for Emergency Core Cooling Systems forLight-Water-Cooled Nuclear Power Plants-Part 50
On December 6, 1978, an advance notice of proposed amendments to Part 50 was published for comment to change certain technical as well as nontechnical requirements within
the existing emergency core cooling system rule.

Codes and Standards for Nuclear Power Plants-Part 50
On December 18, 1978, proposed amendments to Part 50
were published for comment to incorporate by reference new
addenda of a national code that provides rules for the construction of nuclear power plant components.
Indemnification of Spent Reactor Fuel Stored at a Reactor·
Site Different Than the One Where It was Generated-Part140
On January 8, 1979, the Commission published for commenta notice stating its decision to exercise its discretionary
statutory authority under the Price-Anderson Act and extend
Government indemnity to spent reactor fuel stored at a reactor site different than the one where it was generated in two
specific situations. The Commission invited public commenton this decision and on the general policy question raised bysuch an extension.
Codes and Standards for Nuclear Power Plants-Part 50
On January 18, 1978, Proposed amendments to Part 50 were
published for comment to incorporate by reference, withmodifications, a new edition and addenda of the national
code that specifies the requirements for the in-service inspection of nuclear power plant components.
Notices, Instruction, and Reports to Workers: Inspection
Standards for Protection Against Radiation-Parts 19 and 20
On February 20, 1979, proposed amendments to Parts 19 and
20 were published for comment which would eliminate theaccumulated dose averaging formula, 5(N-18), and theassociated Form NRC-4 Exposure History, and impose annual dose-limiting standards while retaining quarterly standards.
Rules of Practice-Part 2
On March 7, 1979, a proposed amendment was published forcomment to amend the rules dealing with ex-parte communications and the separation of adjudicatory and nonadjudicatory functions so that those rules will be in accord
with Government in the Sunshine Act, Pub. L. 94-409.
Domestic Licensing of Production and Utilization
Facilities-Part 50
On March 28, 1979, a proposed amendment to Part 50 was
published for comment specifying fracture toughness andmaterial surveillance program requirements for nuclear reactor to permit greater flexibility in meeting certain of these requirements, and to simplify others by substituting references
to National Standards that have already been incorporated
by reference into the NRC's Regulations.
Financial Protection Requirements and Indemnity
Agreements: Miscellaneow Amendments-Part 140
On April 6, 1979, a proposed amendment to Part 140 was
published for comment to amend regulatiOns relating to thefinancial protection and indemnity required of licensees andto implement legislation that modified and extended for ten
years (to August 1, 1987) the present Price-Anderson legislation.

297

Human Uses of Byproduct Material; Changes in License Conditions for Certain Medical Licenses-Part 35

Safeguards on Nuclear Material; Implementation of
US/IAEA Agreement-Parts 40, 50, 70, 75, 150 and 170

On April 9, 1979, a proposed amendment to Part 35 waspublished for comment making it a requirement to appoint aradiation safety committee that will focus on radiation safety.

On May 1979, proposed amendments were published forcomment to enable the United States to Implement the
US/IAEA Safeguards Agreement, with respect to licensed activities, as soon as the Agreement enters into force.

Public Records-Part 9
On April 17, 1979, an advance notice of proposed rule making was published for comment which would revise the Commission's regulations, "Public Records," implementing theFreedom of Information Act and E.O. 12044 which provides
that regulations be written in "plain English:.
Rules of Practice-Part 2

Adequacy and Acceptance of Emergency Planning Around
Nuclear Facilities-Part 50
On July 17, 1979, an advance notice of proposed rulemaking
was published for comment to establish as conditions of
power reactor operation increased emergency readiness for
public protection in the vicinity of nuclear power reactors on
the part of both the licensee and local and state authorities.

On April 18, 1979, proposed amendments to Part 35 were
published for comment to make procedural changes to permit
the use of new staff procedures on a trial basis for the systemand site safety portions of selected nuclear power plants construction permit applications.

Privacy Act Regulations; Proposed Exemptions-Part 9

Nondiscrimination in Federally Assisted Commission Programs; Application to the Handicapped-Part 4

Packaging of Radioactive Material for Transportation andTransportation of Radioactive Material Under Certain Conditions; Comparability With IAEA Regulations-Part 71

On May 8, 1979, a proposed amendment was published forcomment to implement the requirements of Section 504 of theRehabilitation Act of 1973. The amendment would make it
unlawful for any recipient of Federal financial assistance todiscriminate against a qualified handicapped person, on thebasis of handicap, in employment or the receipt of services.

On August 17, 1979, a proposed amendment was publishedfor comment revising the regulations for the transportation ofradioactive material to make them compatible with those of
the IAEA and thus with those of most major nuclear nationsof the world.

Study of Nuclear Power Plant Construction During Adjudication; Request for Public Comments-Part 2
On June 13, 1979, a notice of proposed rule making was
published for comments seeking the views of the public on theCommission's immediate effectiveness rule (10 CFR 2.764)
which provides that a construction permit can be issued on
the basis of an initial decision of an Atomic Safety and Licensing Board even though that decision is subject to further
review within the Commission.
Testing of Radioisotope Generators-Parts 30 and 35
On June 6, 1979, a proposed amendment was published forcomment to require licensees to test radiopharmaceuticals fora contaminant called molybdenum-99.
Access to and Protection of National Security Information
and Restricted Data-Parts 25 and 95
On July 2, 1979, a proposed amendment was published forcomment governing access to and protection of National
Security Information and Restrictive Data. When former
Atomic Energy Commission regulations were reissued in
March 1975 by the Nuclear Regulatory Commission, rules
governing access to and protection of the National Security
Information and Restrictive Data were not included.

On August 16, 1979, a proposed amendment was publishedfor comment to exempt from certain requirements of thePrivacy Act portions of a proposed new system of records"Special Inquiry File."

Criteria Relating to Uranium Mill Tailings and Constructionof Major Plants-Parts 30, 40, 70, 150 and 170
On August 24, 1979, proposed amendments were publishedfor comment to conform to the requirements of the Uranium
Mill Tailings Radiation Control Act of 1978 and to the standards set forth in the draft generic environmental impact statement on uranium milling.
Privacy Act Regulations; Notices of Proposed Exemptions-Part 9
On August 28, 1979, an amendment was published for comment which proposed exemptions as provided under thePrivacy Act of 1974. The proposed amendment of the Commission's regulations "Public Records" would exempt fromcertain requirements of the Privacy Act portions of a
proposed new system of records "Document Control System."
Production and Utilization Facility Licensees; Emergency
Planning-Part 50 and 70
On September 19, 1979, proposed amendments were
published for comment which would require that all production and utilization facility licensees shall, as a condition of
their license, submit emergency plans for NRC review andapproval and maintain the emergency plans up to date. The
proposed amendments would also require certain special
nuclear material facility licensees (for processing and fuelfabrication, scrap recovery or conversion of uranium hexafluoride) to maintain the emergency plans up to date.

298==================================================
Appendix 5

Regulatory Guides - Fiscal Year 1979
Regulatory guides describe and make available to the
public methods acceptable to the NRC staff for implementing
specific parts of the Commission's regulations and, in somecases, describe techniques used by the staff in evaluating
specific problems or postulated accidents. Guides also may
provide guidance to applicants concerning information need~
ed by the staff in its review of applications for permits andlicenses.
Comments and suggestions for improvements in guides areencouraged at all times, and guides will be revised, as appropriate, to accommodate comments and to reflect new information or experience. In its continuing effort to provide
for increased public participation in the regulatory process,the NRC instituted a new procedure for developing and issuing guides during this fiscal year. Guides are now being issued
for public comment in draft form before the guides havereceived complete staff review and before an official NRC
staff position has been established.
Regulatory guides may also be withdrawn when they aresuperseded by the Commission's regulations, when eqUivalent
recommendations have been incorporated in applicable approved codes and standards, or when changes in methods andtechniques have made them obsolete.
When guides are issued, revised, or withdrawn, notices areplaced in the Federal Register and public announcements
made.
In order to reduce the burden on the taxpayer, the NRC
has made arrangements with the U.S. Government Printing
Office to become a consigned sales agent for certain NRC
publications, Effective November 1, 1979, regulatory guides
are being included in this sales program. Draft guides, which
are issued for public comment, will continue to receive free
distribution. Active guides will be sold on a subscription orindividual copy basis. Licensees of the NRC will receive, at
no cost, pertinent draft and active regulatory guides as they
are issued.
The following guides were issued or revised (or withdrawn
as noted) during the period October 1, 1978 to September 30,
1979.
Division I-Power Reactor Guides
1.7

Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident(Revision 2)

1.9

Selection, Design, and Qualification of DieselGenerator Units Used as Onsite Electric Power
Systems at Nuclear Power Plants (Revision 1)

1.28

Quality Assurance Program Requirements (Design
and Construction) (Revision 2)

1.70

Standard Format and Content of Safety Analysis
Reports for Nuclear Power Plants-L WR Edition(Revision 3)

1.72

Spray Pond Piping Made from Fiberglass-Reinforced
Thermosetting Resin (Revision 2)

1.84

Design and Fabrication Code Case AcceptabilityASME Section III Division 1 (Revisions 14 and 15)

1.85

Materials Code Case Acceptability-ASME Section III
Division 1 (Revisions 14 and 15)

1.104 WITHDRAWN. Overhead Crane Handling Systems
for Nuclear Power Plants
1.125 Physical Models for Design and Operation of
Hydraulic Structures and Systems for Nuclear
Power Plants (Revision 1)
1.128 Installation Design and Installation of Large
Lead Storage Batteries for Nuclear Power Plants(Revision 1)
1.130 Service Limits and Loading Combinations for Class 1
Plate-And-Shell-Type Component Supports (Revision 1)
1.132 Site Investigations for Foundations for Nuclear
Power Plants (Revision 1)
1.134 Medical Evaluation of Nuclear Power Plant
Personnel Requiring Operator Licenses (Revision1)

1.136 Material for Concrete Containment (Revision 1)
1.144 Auditing of Quality Assurance Programs for Nuclear
Power Plants
1.145 Atmospheric Dispersion Models for Potential
Accident Consequence Assessments at Nuclear
Power Plants

Division 2-Research and Test Reactor Guides
2.6

Emergency Planning for Research Reactors

==================================================299
Division 3-Fuels and Materials Facilities Guides

8.20

Applications of Bioassay for 1-125 and 1-131 (Revision
1)

3.11.1 Operational Inspection and Surveillance of Embankment Retention Systems for Uranium Mill
Tailings
3.34

Assumptions Used for Evaluating the Potential
Radiological Consequences of Accidental Nuclear
Criticality in a Uranium Fuel Fabrication Plant(Revision 1)

8.23

Radiation Safety Surveys at Medical Institutions

8.24

Health Physics Surveys During Enriched
Uranium-235 Processing and Fuel Fabrication

Division 9-Antitrust and Financial Review Guides
None

3.35

3.36

3.42

Assumptions Used for Evaluating the Potential
Radiological Consequences of Accidental Nuclear
Criticality in a Plutonium Processing and Fuel
Fabrication Plant (Revision 1)
WITHDRAWN. Nondestructive Examination of
Tubular Products for Use in Fuel Reprocessing
Plants and in Plutonium Processing and Fuel
Fabrication Plants

Division lO-General Guides
10.7

Guide for the Preparation of Applications forLicenses for Laboratory and Industrial Use of
Small Quantities of Byproduct Material (Revision
1)

10.8

Guide for the Preparation of Applications for
Medical Programs

Emergency Planning for Fuel Cycle Facilities andPlants Licensed Under 10 CFR Parts 50 and 70(Revision 1)

3.43

Nuclear Criticality Safety in the Storage ofFissile Materials (Revision 1)

3.44

Standard Format and Content for the Safety Analysis
Report to be Included in a License Application for
the Storage of Spent Fuel in an Independent Spent
Fuel Storage Installation (Water-Basin Type)

Draft Guides
EM 805-5 Nuclear Analysis and Design of Concrete
Radiation Shielding for Nuclear Power Plants
FP 811-4

Safety-Related Permanent Dewatering
Systems for Nuclear Power Plants

MP 711-4

Standard Format and Content for a Licensee
Physical Security Plan for the Protection ofSpecial Nuclear Material of Moderate or Low
Strategic SignificanceDivision 4-Environmental and Siting Guides
4.15

Quality Assurance for Radiological Monitoring
Programs (Normal Operations)-Effluent Streams
and the Environment (Revision 1)
Division 5-Materials and Plant Protection Guides

OH 706-4 Guide for Preparation of Applications for the Useof Gamma Irradiators
OH 714-4 Applications of Bioassay for Fission and Activation Products

5.2

WITHDRAWN. Classification of UnirradiatedPlutonium and Uranium Scrap

OH 717-4 Radiation Protection Training for Light-WaterCooled Nuclear Power Plant Personnel

5.58

Considerations for· Establishing Traceability ofSpecial Nuclear Material Accounting
Measurements

OH 804-4 Audible-Alarm Dosimeters
RH 802-4

Calculational Models for Estimating Radiation
Doses to Man from Airborne Radioactive
Materials Resulting from Uranium Milling
Operations

RS 705·4

Lightning Protection for Nuclear Power Plants

RS 809-5

Qualification Test for Cable Penetration Fire
Stops for Use in Nuclear Power Plants

RS 810~5

Qualification of Quality Assurance Program
Audit Personnel for Nuclear Power Plants

SC 521-4

LWR Core Reloads: Guidance on Applications
for Amendments to Operating Licenses and on
Refueling and Startup Tests

SC 704-5

Functional Specification for Safety-Related
Valve Assemblies in Nuclear Power Plants

Division 6-Product Guides
6.8

Identification Plaque for Irretrievable Well-Logging
Sources
Division 7-Transportation Guides

7.9Standard Format and Content of Part 71 Applications for Approval of Packaging of Type B, Large
Quantity, and Fissile Radioactive Material
Division 8-0ccupational Health Guides

8.19

Occupational Radiation Dose Assessment in LightWater Reactor Power Plants-Design Stage ManRem Estimates (Revision 1)

300

SC 705·4

Ultrasonic Testing of Reactor Vessel Welds
During Inservice Examination

SC 721-4

Inservice Inspection Code Case AcceptabilityASME Section XI Division 1

Testing of Structural Concrete, Structural
Steel, Soils and Foundations During the Construction Phase of Nuclear Power Plants
1.131

Proposed Revision I-Qualification Tests ofElectric Cables and Field Splices for LightWater-Cooled Nuclear Power Plants

5.7

Proposed Revision I-Entry/Exit Control to Protected Areas, Vital Areas, and Material Access
Areas

1.8

Proposed Revision 2-Personnel Selection andTraining

1.33

Proposed Revision 3-Quality Assurance Program Requirements (Operation)

1.35

Proposed Revision 3-Inservice Inspection of
Ungrouted Tendons in Prestressed ConcreteContainments

5.14

Proposed Revision I-Use of Observation(Visual Surveillance) Techniques in Material
Access Areas

1.35.1

Determining Prestressing Forces for Inspectionof Prestressed Concrete Containments

5.44

Proposed Revision 2-Perimeter Intrusion Alarm
Systems

1.58

Proposed Revision I.-Qualification of Nuclear
Power Plant Inspection, Examination, andTesting Personnel

5.57

Proposed Revision I-Shipping and Receiving
Control of Strategic Special Nuclear Material

8.B

Proposed Revision 4-Information Relevant toEnsuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As
Low As Reasonably Achievable (ALARA)

1.94

Proposed Revision 2-Quality Assurance Requirements for Installation, Inspection, and

===================================================301

Appendix 6

Nuclear Electric Generating Units in Operation,
Under Construction or Planned(As of September 30, 1979)

The following listing includes 192 nuclear power reactor electrical generating units which were in operation, under NRC
review for construction permits, and ordered or announced by utilities in the United States at the end of September 1979,
representing a total capacity of approximately 187,000 MWe. TYPE is indicated by: BWR-boiling water reactor,
PWR-pressurized water reactor, HTCR-high temperature gas-cooled reactor, and LMFBR-liquid metal cooled fast breeder
reactor. STATUS is indicated by: OL-has operating license, CP-has construction permit, UR-under review for construction
permit, A/O-announced or ordered by the utility but application for construction not yet docketed by the NRC for review. The
dates for operation are either actual or those scheduled by the utilities (N/S-not yet scheduled).
This listing includes 20 fewer units than a year ago, reflecting cancellations of plans for future facilities. In addition, delays inplanned completion dates have been indicated during fiscal year 1979 for 47 other units. The reasons cited for delays andcancellations include (1) lower demand for electricity, (2) financial problems, (3) construction delays, (4) concerns for reactor
safety, and (5) regulatory delays.

Site

Plant

Capacity(Net MWe)Type

StatusUtility

Commercial
Operation

ALABAMA
Decatur

Browns Ferry Nuclear Power
Plant Unit 1

1,065

BWR

OL 1973

Tennessee Valley Authority

1974

Decatur

Browns Ferry Nuclear Power
Plant Unit 2

1,065

BWR

OL 1974

Tennessee Valley Authority

1975

Decatur

Browns Ferry Nuclear Power
Plant Unit 3

1,065

BWR

OL 1976

Tennessee Valley Authority

1977

DothanJoseph M. Farley Nuclear
Plant Unit 1

829

BWR

OL 1977

Alabama Power Co.

1978

DothanJoseph M. Farley Nuclear
Plant Unit 2

829

PWR

CP 1972

Alabama Power Co.

1980

Scottsboro

Bellefonte Nuclear Plant
Unit 11,235

PWR

CP 1974

Tennessee Valley Authority

1981

Scottsboro

Bellefonte Nuclear Plant
Unit 2

1,235

PWR

CP 1974

Tennessee Valley Authority

1981

302

Plant

Site

Capacity(Net MWe)Type

StatusUtility

Commercial
Operation

ARIZONA
Winterburg

Palo Verde Nuclear Generating
Station Unit 1

1,270

PWR

CP 1976

Arizona Public Service Co.

1982

Winterburg

Palo Verde Nuclear Generating
Station Unit 2

1,270

PWR

CP 1976

Arizona Public Service Co.

1984

Winterburg

Palo Verde Nuclear Generating
Station Unit 3

1,270

PWR

CP 1976

Arizona Public Service Co.

1986

RusselvilleArkansas Nuclear One Unit 1

850

PWR

OL 1974

Arkansas Power & Light Co.

1974

RusselvilleArkansas Nuclear One Unit 2

912

PWR

OL 1978

Arkansas Power & Light Co.

1978

EurekaHumboldt Bay Power Plant
Unit 365

BWR

OL 1962

Pacific Gas & Electric Cu.

1963

San
Clemente

San Onofre Nuclear
Generating Station Unit 1

436

PWR

OL 1967

So. Calif. Ed. & San Diego
Gas & Electric Co.

1968

San
Clemente

San Onofre Nuclear
Generating Station Unit 2

1,140

PWR

CP 1973

So. Calif. Ed. & San Diego
Gas & Electirc Co.

1980

San
Clemente

San On<..fre Nuclear
G~nerating Station Unit 3

1,140

PWR

CP 1973

So. Calif. Ed. & San Diego
Gas & Electric Co.

1981

DiabloCanyonDiablo Canyon Nuclear
Power Plant Unit 1

1,084

PWR

CP 1968

Pacific Gas & Electric Co.

1980

DiabloCanyonDiablo Canyon Nuclear Power
Plant Unit 2

1,106

PWR

CP 1970

Pacific Gas & Elec. Co.

1979

Clay
StationRancho Seco Nuclear
Generating Station Unit 1

917

PWR

OL 1974

Sacramento Municipal Utility
District

1975

Stanislaus Unit 1

1,200

BWR

A/O

Pacific Gas & Elec. Co.

lndef.

Stanislaus Unit 2

1,200

BWR

A/O

Pacific Gas & Elec. Co.

lndef.

Public Service Co. ofColorado

1978

ARKANSASCALIFORNIACOLORADO
Platteville

Fort St. Vrain Nuclear
Generating Station

330

HTGR OL 1973

CONNECTICUT
HaddamNeck

Haddam Neck Generating
Station575

PWR

OL 1967

Conn. Yankee Atomic Power Co.

1968

Waterford

Millstone Nuclear Power
Station Unit 1

660

BWR

OL 1970

Northeast Nuclear Energy Co.

1971. . Site not selected.

303

Site

Plant

Capacity(Net MWe)Type

StatusUtility

Commercial
Operation

CONNECTICUT-Continued
Waterford

Millstone Nuclear Power
Station Unit 2

830

PWR

OL 1975

Northeast Nuclear Energy Co.

1975

Waterford

Millstone Nuclear Power
Station Unit 3

1,159

PWR

CP 1974

Northeast Nuclear Energy Co.

1985

FloridaCity

Turkey Point Station Unit 3

693

PWR

OL 1972

Florida Power & Light Co.

1972

FloridaCity

Turkey Point Station Unit 4

693

PWR

OL 1973

Florida Power & Ligh~ Co.

1973

Red Level

Crystal River Plant Unit 3

825

PWR

OL 1977

Florida Power Corp.

1977

Ft. Pierce

St. Lucie Plant Unit 1

802

PWR

OL 1976

Florida Power & Light Co.

1976

Ft. Pierce

St. Lucie Plant Unit 2

842

PWR

CP 1977

Florida Power & Light Co.

1983

BaxleyEdwin 1. Hatch Plant Unit 1

786

BWR

OL 1974

Georgia Power Co.

1975

BaxleyEdwin 1. Hatch Plant Unit 2

795

BWR

OL 1978

Georgia Power Co.

1978

Waynesboro

Alvin W. Vogtle, Jr. Plant
Unit 11,100

PWR

CP 1974

Georgia Power Co.

1984

Waynesboro

Alvin W. Vogtle, Jr. Plant
Unit 2

1,100

PWR

CP 1974

Georgia Power Co.

1985

MorrisDresden Nuclear Power
Station Unit 1

200

BWR

OL 1959

Commonwealth Edison Co.

1960

MorrisDresden Nuclear Power
Station Unit 2

794

BWR

OL 1969

Commonwealth Edison Co.

1970

MorrisDresden Nuclear Power
Station Unit 3

794

BWR

OL 1971

Commonwealth Edison Co.

1971

Zion

Zion Nuclear Plant Unit 1

1,040

PWR

OL 1973

Commonwealth Edison Co.

1973

Zion

Zion Nuclear Plant Unit 2

1,040

PWR

OL 1973

Commonwealth Edison Co.

1974

Cordova

Quad-Cities Station Unit 1

789

BWR

OL 1972

Comm. Ed. Co.-Iowa-Ill. Gas &Elec. Co.

1973

Cordova

Quad-Cities Station Unit 2

789

BWR

Comm. Ed. Co.-Iowa-Ill. Gas &Elec. Co.

1973

Seneca

LaSalle County Nuclear
Station Unit 1

1,078

BWR

CP 1973

Commonwealth Edison Co.

1979

Seneca

LaSalle County Nuclear
Station Unit 2

1,078

BWR

CP 1973

Commonwealth Edison Co.

1980

FLORIDAGEORGIAILLINOISOL 1972

304

Plant

Site

Capacity(Net MWe)Type

StatusUtility

Commercial
Operation

ILLINOIS-Continued
Byron

Byron Station Unit 1

1,120

PWR

CP 1975

Commonwealth Edison Co.

1982

Byron

Byron Station Unit 2

1,120

PWR

CP 1975

Commonwealth Edison Co.

1983

Braidwood

Braidwood Unit 1

1,120

PWR

CP 1975

Commonwealth Edison Co.

1982

Braidwood

Braidwood Unit 2

1,120

PWR

CP 1975

Commonwealth Edison Co.

1983

Clinton

Clinton Nuclear Power
Plant Unit 1

950

BWR

CP 1976

Illinois Power Co.

1982

Clinton

Clinton Nuclear Power
Plant Unit 2

950

BWR

CP 1976

Illinois Power Co.

1988

Savannah

Carroll County Station Unit 1

1,150

A/O

Commonwealth Edison Co.

1988

Savannah

Carroll County Station Uni t 2

1,150

A/O

Commonwealth Edison Co.

1989

INDIANA
660

BWR

CP 1974

Northern Indiana Public Service
Co.

1984

Marble Hill Unit 1

1,130

PWR

CP 1978

Public Service of Indiana

1982

Marble Hill Unit 2

1,130

PWR

CP 1978

Public Service of Indiana

1984

538

BWR

OL 1974

Iowa Elec. Light & Power Co.

1975

Wolf Creek

1,150

PWR

CP 1977

Kansas Gas & Elec. Co.

1983

Taft

Waterford Steam Electric
Station

1,165

PWR

CP 1974

Louisiana Power & Light Co.

1981

St.
FrancisvilleRiver Bend Station Unit 1

934

BWR

CP 1977

Gulf States Utilities Co.

1984

St.
FrancisvilleRiver Bend Station Unit 2

934

BWR

CP 1977

Gulf States Utilities Co.

NIS

Maine Yankee Atomic Power
Plant

790

PWR

OL 1972

Maine Yankee Atomic Power Co.

1972

Lusby

Calvert Cliffs Nuclear Power
Plant Unit 1

845

PWR

OL 1974

Baltimore Gas & Elec. Co.

1975

Lusby

Calvert Cliffs Nuclear Power
Plant Unit 2

845

PWR

OL 1976

Baltimore Gas & Elec. Co.

1977

Westchester
Town

Bailly Generating Station

MadisonMadison

IOWA
Pala

Duane Arnold Energy Center
Unit 1KANSAS
Burlington

LOUISIANA

MAINE
Wiscasset

MARYLAND

===================================================305
Plant

Site

Capacity(Net MWe)Type

StatusUtility

Commercial
Operation

MASSACHUSETTS
Rowe

Yankee Nuclear Power Station

175

PWR

OL 1960

Yankee Atomic Elec. Co.

1961

Plymount

Pilgrim Station Unit 1

655

BWR

OL 1972

Boston Edison Co.

1972

PlymouthPilgrim Station Unit 2

1,180

PWR

UR

Boston Edison Co.

1985

Turners
Falls

Montague Unit 1

1,150

BWR

UR

Northeast Nuclear Energy Co.

NIS

Turners
Falls

Montague Unit 2

1,150

BWR

UR

Northeast Nuclear Energy Co.

NIA

72

BWR

OL 1962

Consumers Power Co.

1963

805

PWR

OL 1971

Consumers Power Co.

1971

MICHIGAN
Big Rock
Point

Big Rock Point Ntlclear Plant

South Haven Palisades Nuclear Power
StationLagoona
Beach

Enrico Fermi Atomic Power
Plant Unit 2

1,123

BWR

CP 1972

Detroit Power Co.

1980

BridgmanDonald C. Cook Plant Unit 1

1,054

PWR

OL 1974

Indiana & Michigan Elec. Co.

1975

BridgmanDonald C. Cook Plant Unit 2

1,100

PWR

OL 1977

Indiana & Michigan Elec. Co.

1978

MidlandMidland Nuclear Power Plant
Unit 1492

PWR

CP 1972

Consumers Power Co.

1982

MidlandMidland Nuclear Power Plant
Unit 2818

PWR

CP 1972

Consumers Power Co.

1981

St. ClairCountyGreenwood Energy Center
Unit 2

1,200

PWR

UR

Detroit Edison Co.

NIS

St. ClairCountyGreenwood Energy Center
Unit 3

1,200

PWR

UR

Detroit Edison Co.

NISMINNESOTA
MonticelloMonticello Nuclear Generating
Plant

545

BWR

OL 1970

Northern States Power Co.

1971

Red Wing

Prairie Island Nuclear
Generating Plant Unit 1

530

PWR

OL 1973

Northern States Power Co.

1973

Red Wing

Prairie Island Nuclear
Generating Plant Unit 2

530

PWR

OL 1974

Northern States Power Co.

1974

Port Gibson

Grand Gulf Nuclear Station
Unit 1

1,250

BWR

CP 1974

Mississippi Power & Light Co.

1982

Port Gibson

Grand Gulf Nuclear Station
Unit 2

1,250

BWR

CP 1974

Mississippi Power & Light Co.

1984

MISSISSIPPI

306

Site

Plant

Capacity(Net MWe)Type

StatusUtility

Commercial
Operation

MISSISSIPPI -Continued
Yellow
Creek

Yellow Creek Unit 1

1,285

PWR

CP 1978

Tennessee Valley Authority

1985

Yellow
Creek

Yellow Creek Unit 2

1,285

PWR

CP 1978

Tennessee Valley Authority

1991

FultonCallaway Plant Unit 1

1,150

PWR

CP 1976

Union Elec. Co.

1982

FultonCallaway Plant Unit 1

1,150

PWR

CP 1976

Union Elec. Co.

1987

MISSOURI

NEBRASKA
Fort
CalhounFort Calhoun Station Unit 1

457

PWR

OL 1973

Omaha Public Power District

1973

Brownville

Cooper Nuclear Station

778

BWR

OL 1974

Nebraska Public Power District

1974

NEW HAMPSHIRE
Seabrook

Seabrook Nuclear Station Unit 1

1,194

PWR

CP 1976

Public Service of N. H.

1983

Seabrook

Seabrook Nuclear Station Unit 2

1,194

PWR

CP 1976

Public Service of N.H.

1985

Oyster Creek Nuclear Power
Plant Unit 1

650

BWR

OL 1969

Jersey Central Power & Light Co.

1969

Forked River Forked River Generating
Station Unit 1

1,070

PWR

CP 1973

Jersey Central Power & Light Co.

1984

Salem

Salem Nuclear Generating
Station Unit 1

1,090

PWR

OL 1976

Public Service Elec. & Gas Co.

1977

Salem

Salem Nuclear Generating
Station Unit 2

1,115

PWR

CP 1968

Public Service Elec. & Gas Co.

1979

Salem

Hope Creek Generating Station
Unit 1

1,067

BWR

CP 1974

Public Service Elec. & Gas. Co.

1984

Salem

Hope Creek Generating Station
Unit 2

1,067

BWR

CP 1974

Public Service Elec. & Gas Co.

1986

IndianPoint

Indian Point Station Unit 1

265

PWR

OL 19'62

Consolidated Edison Co.

1962

IndianPoint

Indian Point Station Unit 2

873

PWR

OL 1971

Consolidated Edison Co.

1973

IndianPoint

Indian Point Station Unit 3

965

PWR

OL 1975

Consolidated Edison Co.

1976

ScribaNine Mile Point Nuclear Station
Unit 1610

BWR

OL 1969

Niagara Mohawk Power Co.

1969

NEW JERSEY
Toms RiverNEW YORK

===================================================307
Site

Plant

Capacity(Net MWe)

Commercial
Operation

Type

Status

1,080

BWR

CP 1974

Niagara Mohawk Power Co.

1984

Utility

NEW YORK-Continued
ScribaNine Mile Point Nuclear Station
Unit 2'Ontario

R.E. Ginna Nuclear Power
Plant Unit 1

490

PWR

OL 1969

Rochester Gas & Elec. Co.

1970

Brookhaven

Shoreham Nuclear Power
Station854

BWR

CP 1973

Long Island Lighting Co.

1980

ScribaJames A. FitzPatrick Nuclear
Power Plant

821

BWR

OL 1974

Power Authority of State of N.Y.

1975

Long
IslandJamesport Unit 1

1,150

PWR

CP 1979

Long Island Lighting Co.

1988

Long
IslandJamesport Unit 2

1,150

PWR

CP 1979

Long Island Lighting Co.

1990

New Haven 1

1,250

PWR

UR

N.Y. State Elec. & Gas Co.

Indef.

...

New Haven 2

1,250

PWR

UR

N.Y. State Elec. & Gas Co .

Indef.

Sterling

Sterling Power Project Unit 1

1,150

PWR

CP 1977

Rochester Gas & Elec. Co.

1988

NORTH CAROLINASouthport

Brunswick Steam Electric
Plant Unit 2

821

BWR

OL 1974

Carolina Power & Light Co.

1975

Southport

Brunswick Steam Electric
Plant Unit 1

821

BWR

OL 1976

Carolina Power & Light Co.

1977

CowansFord Dam

Wm. B. McGuire Nuclear
Station Unit 1

1,180

PWR

CP 1973

Duke Power Co.

1980

CowansFord Dam

Wm. B. McGuire Nuclear
Station Unit 2

1,180

PWR

CP 1973

Duke Power Co.

1982

BonsaI

Shearon Harris Plant Unit 1

915

PWR

CP 1978

Carolina Power & Light Co.

1983

BonsaI

Shearon Harris Plant Unit 2

915

PWR

CP 1978

Carolina Power & Light Co.

1985

BonsaI

Shearon Harris Plant Unit 3

915

PWR

CP 1978

Carolina Power & Light Co.

1989

BonsaI

Shearon Harris Plant Unit 4

915

PWR

CP 1978

Carolina Power & Light Co.

1987

Davie Co.

Perkins Nuclear Station Unit 1

1,280

PWR

UR

Duke Power Co.

1988

Davie Co.

Perkins Nuclear Station Unit 2

1,280

PWR

UR

Duke Power Co.

1991

Davie Co.

Perkins Nuclear Station Unit 3

1,280

PWR

UR

Duke Power Co.

1993

Davis-Besse Nuclear Power
Station Unit 1

906

PWR

OL 1977

Toledo Edison-Cleveland Elec.
IlIum. Co.

1977

Davis-Besse Nuclear Power
Station Unit 2'" Site not selected.

906

PWR

UR ......

Toledo Edison-Cleveland Elec.
IlIum. Co.

1988

OHIO
Oak Harbor
Oak Harbor

308

Plant

Site

Capacity(Net MWe)Type

StatusUtility

Commercial
Operation

OHIO-Continued
Oak Harbor

Davis-Besse Nuclear Power
Station Unit 3

906

PWR

UR ••

Toledo Edison-Cleveland Elec.
Illum. Co.

1990

Perry

Perry Nuclear Power Plant
Unit 1

1,205

BWR

CP 1977

Cleveland Elec. Illum. Co.

1982

Perry

Perry Nuclear Power Plant
Unit 2

1,205

BWR

CP 1977

Cleveland Elec. Illum Co.

1984

Moscow

Wm. H. Zimmer Nuclear
Power Station Unit 1

810

BWR

CP 1972

Cincinnati Gas & Elec. Co.

1979

Berlin Hgts.

Erie Unit 1

1,260

PWR

UR

Ohio Edison Co.

1989

Berlin Hgts.

Erie Unit 2

1,260

PWR

UR

Ohio Edison Co.

1991

Inola

Black Fox Unit 1

1,150

BWR

UR ••

Public Service Co. of Oklahoma

1983

Inola

Black Fox Unit 2

1,150

BWR

UR ••

Public Service Co. of Oklahoma

1985

PrescottTrojan Nuclear Plant Unit 1

1,130

PWR

OL 1975

Portland General Elec. Co.

1976

Arlington

Pebble Springs Unit 1

1,260

PWR

UR

Portland General Elec. Co.

1986

Arlington

Pebble Springs Unit 2

1,260

PWR

UR

Portland General Elec. Co.

1989

OKLAHOMA

OREGONPENNSYL VANIA
Peach
BottomPeach Bottom Atomic Power
Station Unit 2

1,065

BWR

OL 1973

Philadelphia Elec. Co.

1974

Peach
BottomPeach Bottom Atomic Power
Station Unit 3

1,065

BWR

OL 1974

Philadelphia Elec. Co.

1974

Pottstown

Limerick Generating Station
Unit 1

1,065

BWR

CP 1974

Philadephia Elec. Co.

1983

Pottstown

Limerick Generating Station
Unit 2

1,065

BWR

CP 1974

Philadelphia Elec. Co.

1985

Shippingport Shippingport Atomic Power
Unit 190

PWR

Duquesne Light Co. & DOE

NAShippingport Beaver Valley Power Station
Unit 1852

PWR

OL 1976

Duquesne Light Co.
Ohio Edison Co.

1976

Shippingport Beaver Valley Power Station
Unit 2852

PWR

CP 1974

Duquesne Light Co.
Ohio Edison Co.

1983.... Limited work authorization issued.
1

Operable but OL not required.

309

Plant

Site

Capacity(Net MWe)Type

StatusUtility

Commercial
Operation

PENNSYL VANIA-Continued
Goldsboro

Three Mile Island Nuclear
Station Unit 1

819

PWR

OL 1974

Metropolitan Edison Co.

1974

Goldsboro

Three Mile Island Nuclear
Station Unit 2

906

PWR

OL 1978

Metropolitan Edison Co.

1978

Berwick

Susquehanna Steam Electric
Station Unit 1

1,052

BWR

CP 1973

Pennsylvania Power & Light Co.

1980

Berwick

Susquehanna Steam Electric
Station Unit 2

1,052

BWR

CP 1973

Pennsylvania Power & Light Co.

1982

No. Kingston New England Unit 1

1,150

PWR

UR

New England Power Co.

1987

No. Kingston New England Unit 2

1,150

PWR

UR

New England Power Co.

1989

RHODE ISLAND

SOUTH CAROLINA
Hartsville

H.B. Robinson S.E. Plant Unit 2

700

PWR

OL 1970

Carolina Power & Light Co.

1971

SenecaOconee Nuclear Station Unit 1

887

PWR

OL 1973

Duke Power Co.

1973

SenecaOconee Nuclear Station Unit 2

887

PWR

OL 1973

Duke Power Co.

1974

SenecaOconee Nuclear Station Unit 3

887

PWR

OL 1974Duke Power Co.

1974

Broad River

Virgil C. Summer Nuclear
Station Unit 1

900

PWR

CP 1973

So. Carolina Elec. & Gas Co.

1980

Lake Wylie

Catawba Nuclear Station Unit 1

1,145

PWR

CP 1975

Duke Power Co.

1983

Lake Wylie

Catawba Nuclear Station Unit 2

1,145

PWR

CP 1975

Duke Power Co.

1985

\

')

CherokeeCountyCherokee Nuclear Station Unit 1

1,280

PWR

CP 1977

Duke Power Co.

1986

CherokeeCountyCherokee Nuclear Station Unit 2

1,280

PWR

CP 1977

Duke Power Co.

1988

CherokeeCountyCherokee Nuclear Station Unit 3

1,280

PWR

CP 1977

Duke Power Co.

1988

Daisy

Sequoyah Nuclear Power Plant
Unit 11,140

PWR

CP 1970

Tennessee Valley Authority

1979

Daisy

Sequoyah Nuclear Power Plant
Unit 2

1,140

PWR

CP 1970

Tennessee Valley Authority

1980

SpringCity

Watts Bar Nuclear Plant Unit 1

1,165

PWR

CP 1973

Tennessee Valley Authority

1979

TENNESSEE

310

Site

Plant

Capacity(Net MWe)Type

StatusPWR

CP 1973

Utility

Commercial
Operation

TENNESSEE-Continued
SpringCity

Watts Bar Nuclear Plant Unit 2

1,165

Oak Ridge

Clinch River Breeder Reactor
Plant

350

Hartsville

TVA Plant A Unit 1

Hartsville

TVA Plant A Unit 2

Hartsville

Tennessee Valley Authority

1980

LMFBR UR

U.S. Government

Indef.

BWR

CP 1977

Tennessee Valley Authority

1982

1,205

BWR

CP 1977

Tennessee Valley Authority

1983

TVA Plant B Unit 1

1,205

BWR

CP 1977

Tennessee Valley Authority

1988

Hartsville

TVA Plant B Unit 2

1,205

BWR

CP 1977

Tennessee Valley Authority

1989

PhippsBend

Phipps Bend Unit 1

1,220

BWR

CP 1978

Tennessee Valley Authority

1983

PhippsBend

Phipps Bend Unit 2

1,220

BWR

CP 1978

Tennessee Valley Authority

1986

Glen Rose

Comanche Peak Steam Electric
Station Unit 1

1,150

PWR

CP 1974

Texas P&L, Dallas P&L, Texas
Elec. Service

1981

Glen Rose

Comanche Peak Steam Electric
Station Unit 2

1,150

PWR

CP 1974

Texas P&L, Dallas P&L, Texas
Elec. Service

1983

WallisAllens Creek Unit 1

1,150

BWR

UR

Houston Lighting & Power Co.

1985

Bay CitySouth Texas Nuclear Project
Unit 11,250

PWR

CP 1975

Houston Lighting & Power Co.

1983

Bay CitySouth Texas Nuclear Project
Unit 2

1,250

PWR

CP 1975

Houston Lighting & Power Co.

1985

Vermont Yankee Generating
Station514

BWR

OL 1972

Vermont Yankee Nuclear
Power Corp.

1972

GravelNeck

Surry Power Station Unit 1

822

PWR

OL 1972

Va. Electric & Power Co.

1972

GravelNeck

Surry Power Station Unit 2

822

PWR

OL 1973

Va. Electric & Power Co.

1973

MineralNorth Anna Power Station
Unit 1907

PWR

OL 1976

Va. Electric & Power Co.

1978

MineralNorth Anna Power Station
Unit 2907

PWR

CP 1971

Va. Electric & Power Co.

1979

1,205

TEXAS

VERMONT
VernonVIRGINIA

311

Site

Plant

Capacity(Net MWe)Type

StatusUtility

Commercial
Operation

VIRGINIA-Continued
MineralNorth Anna Power Station
Unit 3907

PWR

CP 1974

Va. Electric & Power Co.

1985

MineralNorth Anna Power Station
Unit 4907

PWR

CP 1974

Va. Electric & Power Co.

1986

Central Virginia 1

1,150

A/O

American Electric Power Co.

1990

Central Virginia 2

1,150

A/O

American Electric Power Co.

1990

WASHINGTON
RichlandN~Reactor/WPPSS Steam

850

GR

RichlandWPPSS No. 1 (Hanford)

1,267

PWR

CP 1975

Wash. Public Power Supply
System1983

RichlandWPPSS No.2 (Hanford)

1,103

BWR

CP 1973

Wash. Public Power Supply
System1981

SatsopWPPSS No.3

1,242

PWR

CP 1978

Wash. Public Power Supply
System1984

RichlandWPPSS No.4

1,267

PWR

CP 1978

Wash. Public Power Supply
System1984

SatsopWPPSS No.5

1,242

PWR

CP 1978

Wash. Public Power Supply
System1985

Sedro
WooleySkagit Nuclear Power Project
Unit 1

1,277

BWR

UR

Puget Sound Power & Light Co.

1985

Sedro
WooleySkagit Nuclear Power Project
Unit 2

1,277

BWR

UR

Puget Sound Power & Light Co.

1987

Wash. Public Power Supply
SystemWISCONSIN
Genoa

Genoa Nuclear Generating
Station (LaCrosse)

50

BWR

OL 1967

Dairyland Power Coop.

1969

Two
CreeksPoint Beach Nuclear Plant
Unit 1497

PWR

OL 1970

Wisconsion Michigan Power Co.

1970

Two
CreeksPoint Beach Nuclear Plant
Unit 2497

PWR

OL 1971

Wisconsin Michigan Power Co.

1972

CarltonKewaunee Nuclear Power
Plant

535

PWR

OL 1973

Wisconsin Elec. Power Co.

1974

Ft.
AtkinsonHaven Nuclear Plant

900

PWR

UR

Wisconsin Elec. Power Co.

1987. . Site not selected.
I

Operable but OL not required.

312

Index
Abnormal Occurrences
Agreement States licensees 175, 176
electrical system deficienci£'.'> 171
emergency feedwater unavailable 172
extortion attempt 139, 171
food contamination 176
loss of containment integrity 75, 171
loss·of·feedwater transient 171
piping reanalysis 91,92, 171
radiography overexposures 176
reports to Congress 175
transport packaging 176
tritium contamination 176
vandalizing new fuel 140, 171
Accident probabilities 240
Accident scenarios 178
Acoustic emission technique 230
Administration & management-see NRC management
Advanced reactors 98·100
Advisory Committee on Medical Uses of Isotopes 131, 288
Advisory Committee on Reactor Safeguards (ACRS) 3, 118·120,286
FFTF review 98
licensing reviews 118
NRC research review 120
personnel, functions 286
TMI·2 accident review 27, 30, 31, 33, 34, 119, 120
unresolved safety issues 120
Adjudicatory proceedings 255-271
Aerosol research 234
Agreement States-see State Agreements program
ALARA (as low as reasonably achievable) implementation 212
Alternate containment concepts 243
Alternate decay-heat removal systems 243
American National Standards Institute (ANSI) 216
Americium-241 130
Anticipated transients without scram (ATWS) 73, 231
Antitrust activiti£'.'> 115, 116
Davis-Besse, Perry decisions 115
review procedure changes 116
South Texas, Comanche Peak reviews 116
St. Lucie plant 116
Stanislaus plant 116
Appellate system study 255
Atmospheric diffusion research 236
Atomic Safety and Licensing Appeal Boards (ASLAB's) 258-260
appellate system study 255
functions, personnel 287, 288
proposal to abolish 255
Atomic Safety and Licensing Boards (ASLB's) 255-258
functions, personnel 285-287
immediate effectiveness rule 255
ATWS-see anticipatoo accidents without scram
Automatic data processing 277
Babcock & Wilcox reactors 19, 25-27
Bioassays 212. Blowdown loading asymmetry 88, 71-73

Breakwaters 102
Breeders 98,99, 234,235
Browns Ferry fire 94
BWR containments 68, 71-73
Civil penalti£'.'> 186-168
Civil Service Reform Act 275
Class 9 accidents 100, 109
Clinch River project 98-see also Breeders
Coal-fired power plants
economics 111, 112
health, safety aspects 100
Coastal zone management plans 105
Code development research 227. 228
Commission appointments 273
Commission decisions 260-262
antitrust decisions 115, 116
fire protection, environmental qualification petition 94
fuel cycle rule 260
seismic reanalysis petition 93
short pilings petition (Bailly plant) 261
Three Mile Island Unit 1 260
uranium fuel cycle review 122, 123
waste disposal, "confidence" rulemaking 261
Compliance-see Enforcement, Inspection
Computer cod£'.'>
assessment 228
development 227
exchange 233
Confidence hearing 147
Congressional hearings 250, 252, 253
Congressional oversight 250
General Accounting Office reports 254
reports to Congress 253, 254
Consolidation of NRC headquarters 279
Consumer products 130, 215, 216
Containment
alternate concepts 243
codes 227-228
design standards 201, 202
regulatory guides 201, 202
structure loading 68, 71·73
sump reliability 85, 86
Contingency planning 139
Control building d£'.'>ign errors 81-88
Control rod guide tube wear 94
Control room simulators 51, 231
Core-disruptive accident 234, 235
Court rulings-see Judicial review
Criticality safety 205
Decommissioning
Babcock and Wilcox Parks Township faCility
financial planning 116, 205
formerly licensed sites 125
Kerr-McGee plutonium fabrication 125
regional workshops 181, 204, 205, 247

125

================================================313
public meetings 24 7
standards 204, 205
state discussions 175, 247
Westinghouse plutonium fabrication 125
Defects, noncompliance reporting 155, 204
Department of Energy (DOE)
Clinch River Breeder Reactor 98
export licensing 186, 187
Federal Radioactive Waste Study 146
gas-cooled reactors 99, 233
high-level waste repository 148, 149
mill tailings, remedial action 152
national laboratories assistance 65
NFS Erwin SNM inventory difference 8, 140
NFS West Valley, N. Y. facility study 125, 126
nuclear safety information coordination 233
research to improve reactor safety 244
spent fuel shipment 137
TMI accident assistance 15, 16, 18
Differing professional opinions (NRC staff) 250
Document control system 278. 279
Domestic safeguards 133-144
Early site reviews 110
Earthquake research 235, 236
ECCS
research 219.223
rulemaking 201
Ecological impact studies 112, 113, 238
Effluent control 100, 101, 239
Electrical connector qualification 90
Electrical equipment qualification 79, 80, 89, 90, 203
Electricity demand forecasting III
Emergency Core Cooling System-see ECCS
Emergency response planning 176-179
accident scenarios 178
costs 179
emergency planning zones 177, 178
GAO report 178
interagency agreements 176, 180, 280
NRC assistance to States 174.178
NRC/DOT Task Force 178
NRC/EPA Task Force 177
overview 5rule change 197
State plans 176, 178, 179
training programs 177, 178transportation. related incidents 129
Employee relations 275
Enforcement 166-168
actions taken 135. 153, 167, 169
civil penalties 167
defects, noncompliance reporting 155, 204
fines imposed 167
goals, initiatives 166, 168
Operations Center 168
orders issued 169
overview 8, 153
types of action 166
Enriched uranium export 190-193-see also Special nuclear material (SNM)
Environmental protection
coastal plants 102
coordination with States 107
effluent control 101

EPA interface 104
exports 194
fuel cycle impact 100, 101
impact statements 110
interagency coordination 104-107
regulatory guides 207, 208
research 238, 239
review process 110
technical specificatiOns 101
social, economic issues 110·112
standards 207, 208
Environmental qualification, see Electrical equipment qualification
Environmental Standard Review Plans 110
EPICOR-II 13, 23, 24
Equal Employment Opportunity Program 276, 277
Export-import activities 9, 186-194
Export licensing 189-195
environmental effects 194
health and safety implications 9, 189
interagency coordination 186, 187
interventions 189, 193
Nuclear Nonproliferation Act 186, 187
Philippines reactor 9, 189, 193
power reactors 9, 189, 193
safeguards review 194
significant cases 189, 193
source materials 192
special nuclear material 190-192
tabulation 190-193
Tarapur (India) case 193, 194
Extraordinary Nuclear Occurrence (ENO) 118
Fast Flux Test Facility (FFTF) 98, 118, 119, 234, 235
Feedwater nozzle cracking 73-75, 88, 89
Federal Emergency Management Agency (FEMA)
role in nuclear emergencies 5, 61, 176, 280
Financial statements (NRC) 281,282
Fish impingement, entrainment 101, 102, 112
Fire protection 204, 231
Flaw detection 230
Floating nuclear power plants 99, 100, 102
Flood research 236
Floodplain management 102
Fort St. Vrain reactor 99, 233
Freedom of Information Act 246, 247
Fuel behavior research 224-227
Fuel cladding research 224
Fuel cycle
criticality safety 205
decommissioning 125, 126, 204
environmental survey 122, 123
EPA uranium cycle standard 121
fuel fabrication 212
milling 151, 152
plant safety 205, 206
plutonium processing 124, 125
reprocessing-recyle 125, 126
research 238, 239
resident inspectors 6, 153, 154, 159
risk assessment 241
Table S-3 rulemaking 122, 123, 260
tailings management 151, 152
Fuel meltdown research 226
Funding and budget 278-280

314

Gamma irrapiators 213
Gas.cooled reactors 99, 233
General Accounting Office (GAO) reports 3, 178,254
General Electric Test Reactor (GETR) 94
Geology, seismology research 235, 236
Geotechnical and sciences (waste) program 239
Government in Sunshine Act 248, 249
Greene County (N.Y.) Site 101
Halden reactor test codes 226, 227
Health effects
coal, nuclear 100
low·level radiation 208, 209
uranium milling 212
High-level waste management 7, 145, 147-149
High-temperature gas· cooled reactor (HTGR) 99, 233
Human factors research 231-see also TMI-2 accident
Humboldt Bay site 93
Hydrogen bubble concern 17-1Q
Hydrology research 236
IAEA-see International Atomic Energy Agency
IE Bulletins, Circulars, Information Notices 159-166
Immediate effectiveness rule 255
Improved in-plant accident response 243
Improved reactor safety research 242-244
Incident Response Program 153, 168
Indemnity financial protection 116.118
Independent Spent Fuel Storage Facility 124
In situ solution mining i51
Industrial radiography-see Radiography
Inspection 8, 153.166
bulletins, circulars, information- notices 159·166
construction 155, 166
goals, initiatives 153, 155, 157
Incident Response Program 153, 168
Independent measurement verification 157
licensee quality assurance 155
NRC Staff 153-155
performance appraisal 157
power reactors 155-157
reactive inspections 166
reporting defects, noncompliance 155, 204
resident inspection program 159
resident inspectors (tabulation) 154
safeguards 155
TMI-2 accident 157-159
types 155
vendors 155.157
workload 155-see also Enforcement
Inspector & Auditor (IA) 277, 278
Insurance premium refunds 118
Integral systems tests 220-222
Intergovernmental personnel assignment 180, 181
International activities 183-196
bilateral arrangements 183, 184
cooperation with IAEA, OECD 216, 217, 185, 186
exports, imports 186-194
foreign visitors 185

information exchange 183-186
multinational projects 184, 185, 221
nuclear fuel cycle evaluation 187, 188
reactor exports 189, 193, 194
research agreements 184, 185safeguards 194-196
technical exchanges 186
3D program, research 222,223
International Atomic Energy Agency (IAEA)
reactor safety standards 185, 186, 216
safeguards agreement 142, 194, 195
safety assistance 185, 186
technical assistance 186transportation standards 129
International Nuclear Fuel Cycle Evaluation (INFCE)
Investigations by IA 277
Investigations by IE 168-172

98, 187, 188

Judicial review 262·271
cases concluded 264-267
cases initiated 267-268
cases pending 268·271
significant cases 262-264
Kemeny Commission-see President's Commission
Legislation
civil penalties amendment 168
mill tailings 151
NRC Authorization Act (1979) 242
State participation in waste management 147
Lewis Risk Assessment Review Group 240. License fees 279
License revocation, suspension-see Enforcement
Licensing processdescription (reactors) 64
improving procedures 107-115
Liquid effluent dispersion 239
Liquid Pathway Generic Study 99, 100
Liquid display watches 130
Liquid Metal Fast Breeder Reactor (LMFBR) 234, 235
Litigation-see Judicial review_Load com binations study 237
LOCA-see Loss-of-coolant accident
LOFT-see Loss·of·Fluid Test Facility
Loss-of-coolant accident (LOCA) codes 227
Loss·of·Fluid Test Facility 184, 219-226
Low-level radiation effects 199, 200, 208, 238
Low-level waste disposal 7,126, 127, 149-152,206, 240
LWR fuel cycle-see Fuel cycle
Management, administration-see NRC management
Manufacturing license 97-100
Materials control, accounting 134, 142, 143, 215
Materials regulation...-see Nuclear materials
Materials research . 228·230
Materials (waste) science program 239
Mechanical engineering research 236
Medical policy statement 209, 210

315

Meltdown research 226
Metallurgy research 228
Meteorology studies 236
Mill tailings-see Uranium milling
Minorities 276
National Energy Software Center 233
National standards program 216
National Treasury Employees Union 275
Natural phenomena 235, 236
Need for power study 111
NEPA review-see Environmental protection
Neutron dosimetry 211
New staff offices 275
Noncompliance reporting 155, 170, 204
Nonpower reactor safeguards 140
Nonproliferation Alternative Systems Assessment Program
98, 187, 188-see also International Nuclear Fuel Cycle Evaluation
Nonproliferation policy 186·189
NRC management
automated document control 278
automatic data processing 277
committees, boards 286·288
contracting and reimbursable work 278
equal employment opportunity 276
financial statements 281, 282
funding 278·280
inspection and audit 277
license fees 279
minority employment 275
national emergency preparedness 280
Operations Center 153, 168
organization 273, 274, 283-285
Overview 1-10
personnel 273
physical facilities 279
public document rooms 289
NRC policies, goals 1, 2, 9, 10
NRC/EPA memorandum of understanding 179
NRC/DOT coordination 178,Nuclear fuel cycle evaluation 187, 188
Nuclear materials
regulation 121-132
transport 126-130-see also Radioactive wastes, Special nuclear material
Nuclear medicine 131, 132
Nuclear Nonproliferation Act of 1978
NRC views 188
Nuclear Plant Reliability Data System (NPRDS) 242
Nuclear power reactors-see Power reactors
Nuclear Regulatory Commission Issuances 258
Nuclear Safety Information Center 233
Occupational exposures 210-213
fuel fabricators 212
radiographers 213
reactor personnel 210, 211
Occupational health standards 104, 210-213
Operating experienceabnormal occurrences 91, 92, 139, 140, 171-176
NRC organization 275

occupational exposures 104, 210, 211-see also Systematic Evaluation Program
Operator licenses 65, 204
Organizaton for Economic Cooperation and Development-see also International Activities
Overpressurization 80
Overview of report 1-10
Packaging standards 127-129
PARET program 236
Pennsylvania Emergency Management Administration 15
Personnel (NRC)
appointments 273
Civil Service Refonn Act 275
equal employment opportunity 276
headquarters consolidation 3, 279
Inspector and Auditor investigations 277
major changes 273
numbers, professions 273
organizational changes 273
policy on differing professional opinions 250
union activity 275
Personnel (licensee) clearance 138
Philippines Reactor Project 9, 189, 193
Physical security-see Safeguards
Pipe cracks 82-85, 88, 89
Piping benchmarks 237
Plutonium
facilities decommissioning 125
packaging 129
processing, fabrication 124, 125
Policy, Planning and Program Guidance 1, 2Power Burst Facility 225
Power reactors
abnormal occurrences 171
accident consequences 241
accident scenarios 178
advanced plants 98-100
advanced reactor research 233·235
alternative sites 102-104, 108
anticipated transients without scram (ATWS) 73, 231
antitrust reviews 115, 116
blowdown loading 68, 71-73
building settlement 170
bulletins, circulars, information notices 159-166
breeders 98
BWR nozzle cracking 73-75
civil penalties 167
concrete deficiencies 169, 170
comparison with coal-fired plants 100, 111, 112construction permits 63, 65, 301-311
containment sump reliability 85, 86
control building design 86-88
control rod guide tubes 74,94
coolant pump supports 76, 77
cooperation with states 107
decommissioning 125, 204, 205
degraded core cooling 197
effluent control 100, 101
electrical equipment qualification 79, 80, 89, 90, 171
emergency feedwater unavailable 172
enforcement orders 169
environmental protection 104-107
environmental review 100·104, 108-113
exports 189, 193"Extraordinary Nuclear Occurrence" 118

184, 185

316

Fast Flux Test Facility 98, 118, 119, 235
feedwater nozzle cracks 73-75, 88, 89
financial protection 116-118
fire protection 94
floating plants 99, 100
fuel cask drop 170
fuel rod vandalizing 171
gas-cooled reactors 99, 233
generic design deficiencies 93
health effects 100
in~plant accident response 243
inspections 157-159
installation nonconformance 90 91
interagency coordination 104-uh~ntern~tio~al safety standards 185, 186, 189, 216, 217
Investigations 169-171
licensing process 64, 107-115
licensing status 63, 65
loose parts detection 231
loss of containment integrity 89, 171
loss of electric power 86
loss-of-feedwater transient 171
manufacturing license 97-100
materials, components standards 201, 202
need-for-facility issue III
occupational exposure 104, 210, 211
offsite explosion protection 235
operating experience evaluation 5, 6, 275
operating licenses 63, 65, 301-311
operator licenses 65, 204pipe cracks 82-85, 88, 89
pipe support loading 90
piping system installation 90, 91
pressure suppression containments 71-73
pressure transients 80
purge-valve operability 89
quality assurance 90, 91, 113, 114, 202
qualification tests 79, SO, 203radiation safety training 213
radiological impact 100, 101
reactor vessel integrity 75, 76, SO
regulatory priorities 1, 2
research to improve safety 242-244
resident inspectors 6, 153, 154, 159, 277
resi~ual heat removal (post-shutdown) SO, 81
respuatory protection 210
risk assessment 66
rulemaking 107, 108
safeguarding 138, 139, 144
safety overview 3, 4
seismic design 82. 191-194
simulation 198
siting policy 102-104, 108-110
siting problems 101·104
socioeconomic impacts 110-112
spent fuel storage 81, 82, 117. 118
standard plants 95-98
standard review plans 110
standard technical specificatiOns 101
standards 200·204
station blackout 86
status 63
steam generator supports 76, 77
steam generator tube integrity 69, 70, 230
Systematic Evaluation Program 81, 114, 115
system interaction 77
tabulations 63, 65, 96, 97, 301·311
technical problems 86-95
technical specifications (environmental) 102
unresolved safety issues 65-86
water hammer 67, 68, 237

water reactor research 219-233
President's Commission (TMl-2 accident)
findings, judgments 35-39
NRC responses 42-60
President's response 60-62
recommendations 42-62
supplemental views 39-42
Pressure vessel research 229
Price-Anderson Act 116-118
Privacy Act 246, 247
Procurement review 277
Public Participation
Congressional hearings 250, 252, 253
enhan(,,'ement 250
formal participation 249, 251
Government in Sunshine Act 248
informal participation 247. 248
NRC information program 245.247
public document rooms 245, 246, 289-293
TMI·2 accident 246
PWR feedwater line cracking 88, 89
Qualification testing 231
Quality assurance 90, 91, 113, 114, 202
Radiation dosimetry research 238
Radiation exposure
consumer products 215
low.level, health effects 208, 238
occupational 104, 199, 200, 210-213
personnel dosimetry 210, 238"Radiation Area" petition 213
radiographers 213
respiratory protection 210, 212
teletherapy 132
TMI-2 accident 20.22, 36transportation 128
uranium fuel cycle 211, 212
Radiation protection policy 6Radioactive materials-see Nuclear materials
Radioactive wastes
burial site surveys 125
confidence hearing 147
decommissioning 125, 126,150, 204,205
DOE program 148, 149
environmental impact statements 147, 149
formerly licensed sites 125
high-level 147-149
Interagency Review Group 146
licensing 148, 149-151
low-level 149-152
mill tailings 151, 152
NRC organization 145
overview 7, 145
policy 1, 2
regulation 145, 148.151, 206
rulemaking, Commission confidence 261
repositories 145, 148, 149
research 239, 240
siting 145, 151
standards guides 206
State participation study 146, 147
TMI·2 accident 150
transport 126, 127
uranium recovery 151, 152
Radiographyincidents 176

317

industrial 213
licensing 173
Radioisotopes licensing 130-132
consumer products 130
industrial 130, 131
medical 131, 132
Radiological emergencies-see Emergency Response Planning
Radon-222 122, 123
Reactor regulation 63-120
Reactors, see Power reactors
Reactors Safety Study, review of 240
Reactor vessel integrity 75, 76, 80
Regulations, amendments (FY 1979) 294·297
Regulatory guides, standards 197-217, 298-300
consumer products 215, 216
containment design, structures 201, 202
current priorities 197-200
environmental protection 207, 208
emergency planning 176, 197, 206, 207
fire protection 204
fuel cycle facilities 204·206
high-level waste 147
international participation 216, 217
issuances of guides in FY 1979 298-300
low-level waste 149
national program 216
occupational health 199, 200, 210-213
packaging 127, 129
power reactors 200-204
pressure vessel 201
qualification testing 203
radioisotopes in industry 215, 216
radioisotopes in medicine 209, 211
safeguards 213, 215
seismic design 207
siting 206, 207transportation 200
waste management 205,206
Reorganization recommendations 274
Reprocessing-recycle issue
international aspects 186-188
Resident inspection program 6, 154, 159, 277
Research
acoustic monitoring 230
advanced reactor research 233
aerosol release and transport 234
alternate containment concepts 243
alternate decay heat systems 243
cladding research 244
code development/assessment 227, 228
counter current flow limit 222, 223
cylindrical core test facility (Japan) 222
debris-bed behavior 235
ecological impacts 238
environmental research 238, 239
Fast-Flux Test Facility 235
fire protection research 231
fission-product release, transport 226
flaw detection 230
FLECHT-SEASET program 222
fracture mechanics 229
fuel behavior 224
fuel cycle research 238
fuel meltdown 226
gas-cooled reactors 233
geology and seismology 235
geotechnical and sciences program 239

high-level waste research 239
human factors research 231
hydrology (and meteorology) 236
information centers 233
improved accident response 243
improved safety for nuclear power plants 242
instrument development 223
integral systems tests 220
international agreements 184, 185
LMFBR research 234
LOFT experiments 220
low-level waste research 240
material science programs 239
mechanical engineering 236
meltdown studies 226
metallurgy, materials 228
materials interactions 234
meteorology and hydrology 236
Multirod Burst Test program 224
noise diagnostics 231
operational safety research 231
overview 6, 7
policy 2
Power-Burst Facility 225
pressure vessel integrity 229
PWR blowndown heat transfer 222
qualification testing 231
radiation dosimetry 238
radiation embrittlement 230
reactor safety improvement 242
research support 230
risk assessment 240
safety, relief valves 231
seismic studies 237
seismology (and geology) 235
semiscale program 221, 277
separate effects experiments 222
site safety 235
steam generator tubes 230
steam-water mixing tests 220
stress corrosion cracking 230
structural engineering 237
systems engineering 219
three-dimensional (3D) flow programs 227, 233
tornadoes 238
transportation 238
tubing 230
two-loop test apparatus 222
Upper Plenum Test Facility (Germany) 222
uranium mill research 240
W ASH-14oo review (Lewis Group) 240
waste management 239
water hammer 237
water reactor safety 219-224, 233
Zircaloy research 224, 225
Respiratory protection 210
Risk assessment 1, 6, 66, 240-242
ACRS review 120
data analysis 242
fuel cycle 241
licensing support 241
methodology development 241, 242
research 240
Review Group 240
unresolved safety issues 66, 241
Safeguards, domestic 8, 133-144
adequacy 133, 134
contingency planning 139, 140
effectiveness evaluation methods
fuel cycle facilities 134, 135

142-144318

incidents 139, 140
inspections 136, 139
Integrated Safeguards Program Plan 144
interagency coordination 139
material accounting 142, 215
nonpower reactors 140
NRC management 144
personnel clearances 138
physical protection equipment 213, 215
physical security upgrade rule 8, 135, 277
reactor sabotage protection 138, 139, 142-144
report (P.L. 95-601) 133-144
regulations 141, 142
research 142-144
sabotage attempt 140
standards 213, 21Stransportation 13S-139
upgrading 140, 141
Safeguards, international 9, 142, 144, 193, 194-196
Safety policy 1, 2, 10
Seismic design criteria 93
Seismic problems
General Electric Test Reactor 94
Humboldt Bay 93
interagency coordination 106
Trojan Nuclear Plant 86-88
reanalysis of piping systems 91, 92
Seismic safety margins 236
Seismology research 235, 236
Semiscale program 221
Senior Executive Service Program 275
Separate effects experiments 222-224
Severe weather pheonomena 236
Siting of facilities
alternative sites 102-104, 108
coastal 102
coordination with States 107, 207interagency coordination 104-107
nearby explosive hazards 109
problems 101-104
regional siting 239
research 235, 236
standards 206, 207
Siting Policy Task Force 108-110
SNM-see Special nuclear material
Socioeconomic impacts 239
Source material export 192
Special nuclear material (SNM)
accounting 142, 21S
export 190-192
inventory discrepancies 8, 135, 140, 21S
safeguarding 140, 141, 194-196transport security 141
Spent reactor fuelGElS 123
indemnification issue 117
licensing 123, 124
overhead crane control 81, 82
Pacific Basin storage 188
reprocessing-recycle issue 187, 188
storage 188, 20S
transport 8, 9, 128, 129, 135-137
SSNM-see Strategic special nuclear material
Standard Review Plans

110

Standard technical specifications 101, 102
Standardization of reactors 9S-98
Standards-see Regulatory guides, standards
State Agreements Program 173·17S
abnormal occurrences 175, 176
adequacy, compatibility findings 173
annual meeting 175
low-level waste disposal 14S, lSI
milling cleanup 151, 152
NRC annual review 173
radiation control program 17S
radioactive waste management 146, 147
technical assistance 151, lS2, 174
training State personnel 174
uranium recovery licensing lSI, 17S
Statescoastal zone management lOS
emergency response planning 176·179
environmental review 107
intergovernmental personnel assignments 180, 181
legislation 8, 180, 247
liaison officers 179, 180
licensing coordination 179
memoranda of understanding 179
radiation control programs 180
regional workshops 181
siting coordination 206, 207
technical assistance 174-178
transportation surveillance 178
Station blackout 86
Steam generator tube integrity 69, 70, 230
Steam-water mixing tests 222
Steel piping cracks 82-85, 88, 89
Strategic special nuclear material (SSNM)
export 189
safeguarding 133-144
Stress corrosion 230
Summary of Annual Report 1-10
Susquehanna River monitoring 21, 22
Systematic Evaluation Program 81, 114, lIS
Systems engineering research 219-224
Tailings-see Uranium milling
Tarapur (India) export 193, 194
Three Mile Island (TMI) Nuclear Station
description 11
photos 12, 13
TMI Unit 1 shutdown 260
TMI-Vnit 2
control room 18
schematic 14, 15
TMI-Vnit 2 accident
ACRS reviews 27,30, 31, 33, 34
action plan 31
BNL assistance 16
Bulletins & Orders Task Force 24-27
bulletins (NRC) issued 19, 25, 27, 161
claims handling procedures 117, 118
clean up 22-24, 32, 40
communications facilities IS, 17. 29, 247
Congressional concern 17
contaminated water treatment 23, 24
DOE assistance IS, 16, 18
effect on NRC inspections 158, lS9effect on NRC policy 1, 2

===================================================319
effluent treatment 101
emergency response planning 33, 37, 38
environmental protection 21, 22
Epicor II 13, 23, 24
Extraordinary Nuclear Occurrence determination 118
Federal agencies' assistance IS, 18
financial protection 116, 117
human factors involved 29-31, 34-36, 38-40, 48-51
hydrogen bubble concern 17-19hydrogen burn 16, 17
impact on inspection program 158
individual doses 20
Inspection & Enforcement Task Force 32-34
inspections resulting 157, 158
investigation center 246
investigation reports 11, 278
Kemeny Commission-see President's Commission
Lessons Learned Task Force 27-31
limited evacuation advisory 17-19
major components involved 14, 15
meterological measurements 20, 21
narrative 11-20
NRC Operations Center 15, 16
NRC orders on B&W plants 19,25-27
NRC response to President's Commission 42-60
NRC Special Inquiry Group 3, 11, 20
occupational doses 20
on-site NRC staff 16, 18, 23
population doses 20
popldation evacuation 17-19
President's Commission investigation 2, 35-62
President's response to PC investigation 60-62
public information centers 247
radiation monitoring 15, 16, 19-22
radiological consequences 20-22, 36
recovery operations 22·24, 32, 40
related generic studies 26
States' assistance 178
socioeconomic impacts III
Susquehanna river monitoring 21, 22
telephone survey III
waste shipments 150, 180, 181
Tornado research 238
Transient worker protection 211
Transport of nuclear materials
accidents, emergency plans 178
DOT/NRC surveillance 178
emergency response planning 129
Federal coordination 126, 129
lAEA standards 129, 130
irradiated fuel 123, 127, 128
low-level wastes 126, 127
overview 8
packaging 127-129
radioactive wastes 150, 151
safeguarding 141*U.s. GQVERNMENT PRINTING OFFICE: 1980

620-269/171

1-3

State liaison 179
urban areas 128, 129
worker protection 128
Tritium contamination 176
Trojan Nuclear Plant 86-88
Two-loop Test Apparatus 222
Union activity 275
Unresolved Safety Issues 65-86
evaluation process 66
new issues 66
progress reports 67
tabulation 67-see also Power reactors
Uranium, depleted 192
Uranium, enriched
export 190-192
safeguarding 193, 194-196
Uranium enrichment 192, 193
Uranium export 192
Uranium fuel cycle-see Fuel cycle
Uranium millingbioassay 212
DOE remedial actions 152
GElS 151
health protection 205, 212
impoundment failure 146, 152
in situ leaching 151
licensing reviews 151
radiation levels 212
research 240
tailings 205, lSI, 152
Uranium Mill Tailings Control Act of 1978 151, 152, 175, 205
Uranium mining 151
Vendor inspections 155-157
Violations-see Enforcement
WASH-1400-see Reactor Safety Study
Waste Isolation Pilot Program (WIPP) 148
Waste management-see Radioactive wastes
Waste heat utilization 106
Water hammer 67,68,237
Water reactors-see Power reactors
Water reactor research 219-233
Well logging 131
West Valley (N.Y.) Plant 125, 126
Women's programs 276
Zircaloy research

224, 225

NUREG·0690