NUREG-1537, Part 1

Guidelines for Preparing and 

Reviewing Applications for the 

Licensing of Non-Power Reactors 

Format and Content 

Office of Nuclear Reactor Regulation 
NUREG-1537, Part 1 

Guidelines for Preparing and 

Reviewing Applications for the 

Licensing of Non-Power Reactors 

Format and Content 

Manuscript Completed: February 1996 

Date Published: February 1996 

Office of Nuclear Reactor Regulation 
UNITED STATES

NUCLEAR REGULATORY COMMISSION

s WASHINGTON, D.C. 20555-001

March 6, 1996

TO: ALL HOLDERS OF OPERATING LICENSES OR CONSTRUCTION PERMITS FOR TEST

AND RESEARCH REACTORS

SUBJECT: ISSUANCE OF NUREG-1537, "GUIDELINES FOR PREPARING AND REVIEWING

APPLICATIONS FOR THE LICENSING OF NON-POWER REACTORS"

The U.S. Nuclear Regulatory Commission (NRC) has issued NUREG-1537,

"Guidelines for Preparing and Reviewing Applications for the Licensing'of

Non-Power Reactors." Part 1 of NUREG-1537 contains format and content

guidance for non-power reactor (NPR) applicants and licensees and Part 2

contains a standard review plan and acceptance criteria for NRC NPR

reviewers.

The format and content guide suggests a uniform format for presenting

information in NPR applications, helps ensure completeness of information

provided, assists the Commission staff and others in locating information, and

aids in increasing the efficiency of the review process. The format and

content guide represents a format for NPR applications that is acceptable to

the NRC staff. Conformance with the format and content, however, is not

required.

The standard review plan ensures the quality and uniformity of the staff

reviews, makes information about regulatory matters concerning NPRs widely

available, and improves the understanding of the staff review process by

interested members of the NPR community and the public.

The document covers all aspects of NPR licensing. The document can be used

for the construction permit and the initial operating license, license

renewal, license amendment, decommissioning and license termination, and

highly enriched to low-enriched uranium core conversions. There is also an

appendix to the format and content guide that lists selected regulations that

are applicable to NPRs.

The document chapters were released in draft form for public comment as the

staff completed them. The staff evaluated comments from interested parties

and a number of them were incorporated into the document. The comments did

not result in any major changes to the document, but the staff used them to

clarify the documents. The draft documents were changed in response to

comments and the documents were edited to make the writing style consistent

between chapters. The staff responded to everyone who commented on the draft

documents by sending commenters the staff analysis of their comments and

pointing out any changes made to the text as a result of the comments.

Comments and NRC responses were placed in the Public Document Room.
- 2 -

The document is in looseleaf form to be put in three-hole binders for ease of

use. We plan to amend the document as necessary to keep its content current.

Your comments on NUREG-1537 are encouraged. They should be sent to--

Director, Non-Power Reactors and Decommissioning

Project Directorate

United States Nuclear Regulatory Commission

M. S. 0-1l B-20

Washington, DC 20555-0001

Comments will be considered in future revisions of the document. Questions

concerning this project should be directed to the project manager for this

effort, Alexander Adams, Jr., at 301-415-1127.

Seymour H. Weiss, Director

Non-Power Reactors and Decommissioning

Project Directorate

Division of Reactor Program Management

Office of Nuclear Reactor Regulation
u-rwvr net Owc1

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--.. Office or -Nucle'ar Reactof-'r:T7 ...... Division 6Reaco rga Manameni

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AVAILABILITY NOTICE

) Availability of Reference Matenals Cited in NRC Publications

Most documents cited in NRC publications will be available from one of the following sources:

1. The NRC Public Document Room. 2120 L Street. NW.. Lower Level, Washington, DC

20555-0001

2. The Superintendent of Documents. U.S. Government Printing Office. P. 0. Box 37082.

Washington, DC 20402-9328

3. The National Technical Information Service. Springfield, VA 22161-0002

Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Public

Document Room include NRC correspondence and internal NRC memoranda; NRC bulletins,

circulars, information notices. inspection and investigation notices: licensee event reports;

vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.

The following documents in the NUREG series are available for purchase from the Government

Printing Office: formal NRC staff and contractor reports. NRC-sponsored conference proceedings, international agreement reports, grantee reports, and NRC booklets and brochures. Also available are regulatory guides. NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances.

Documents available from the National Technical Information Service include NUREG-senes

reports and technical reports prepared by other Federal agencies and reports prepared by the

Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

Documents available from public and special technical libranes include all open literature

items. such as books, journal articles, and transactions. Federal Register notices, Federal

and State legislation, and congressional reports can usually be obtained from these libranes.

Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.

Single copies of NRC draft reports are available free, to the extent of supply, upon written

request to the Office of Administration. Distribution and Mail Services Section, U.S. Nuclear

Regulatory Commission, Washington DC 20555-0001.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory

process are maintained at the NRC Library. Two White Flint North.1 1545 Rockville Pike, Rockville. MD 20852-2738. for use by the public. Codes and standards are usually copyrighted

and may be purchased from the originating organization or, if they are American National

Standards, from the American National Standards Institute. 1430 Broadway, New York. NY

10018-3308.

Photograph on cover courtesy of General Atomics
UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-o00

March 6, 1996

TO: ALL HOLDERS OF OPERATING LICENSES OR CONSTRUCTION PERMITS FOR TEST

AND RESEARCH REACTORS

SUBJECT: ISSUANCE OF NUREG-1537, "GUIDELINES FOR PREPARING AND REVIEWING

APPLICATIONS FOR THE LICENSING OF NON-POWER REACTORS"'

The U.S. Nuclear Regulatory Commission (NRC) has issued NUREG-1537,

"Guidelines for Preparing and Reviewing Applications for the'Licensing of'

Non-Power Reactors." Part 1 of NUREG-1537 contains format and content

guidance for non-power reactor (NPR) applicants and licensees and Part 2

contains a standard review plan and acceptance'criteria for NRC NPR

reviewers.

The format rind content guide suggests a uniform format for presenting

information in NPR applications, helps ensure completeness of information

provided, assists the Commission staff.'and others in locating information, and

aids in increasing the efficiency of the review process. The format and

content guide represents a format for NPR applications that is acceptable to

the NRC staff. Conformance with the format and content, however, is not

required.

The standard review plan ensures the quality and uniformity of the staff

reviews, makes information about regulatory matters concerning NPRs widely

available, and improves the understanding of the staff review process by

interested members of the NPR community and the public.

The document covers all aspects of NPR licensing. The document can be used

for the construction permit and the initial operating license, license

renewal, license amendment, decommissioning and license termination, and

highly enriched to low-enriched uranium core conversions. There is also an

appendix to the format and content guide that lists selected regulations that

are applicable to NPRs.

The document chapters were released in draft form for public comment as the

staff completed them. The staff evaluated comments from interested parties

and a number of them were incorporated into the document. The comments did

not result in any major changes to the document, but the staff used them to

clarify the documents. The draft documents were changed in response to

comments and the documents were edited to make the writing style consistent

between chapters. The staff responded to everyone who commented on the draft

documents by sending commenters the staff analysis of their comments and

pointing out any changes made to the text as a result of the comments.

Comments and NRC responses were placed in the Public Document Room.
- 2 -

The document is in looseleaf form to be put in three-hole binders for ease of

use. We plan to amend the document as necessary to keep its content current.

Your comments on NUREG-1537 are encouraged. They should be sent to--

Director, Non-Power Reactors and Decommissioning

Project Directorate

United States Nuclear Regulatory Commission

M. S. 0-11 B-20

Washington, DC 20555-0001

Comments will be considered in future revisions of the document. Questions

concerning this project should be directed to the project manager for this

effort, Alexander Adams, Jr., at 301-415-1127.

Seymour H. Weiss, Director

Non-Power Reactors and Decommissioning

Project Directorate

Division of Reactor Program Management

Office of Nuclear Reactor Regulation
ABSTRACT

NUMREG-1537, Part I gives guidance to non-power reactor licensees and

applicants on the format and content of applications to the Nuclear Regulatory

Commission for licensing actions. These licensing actions include construction

permits and initial operating licenses, license renewals, amendments, conversions

from highly enriched uranium to low-enriched uranium, decommissioning, and

license termination.

REv 0,2/96 iii STANrWW FORMAT AND Cowrr

REV 0, 2/96 . STANDARD FoRmAT AND CoNTENT
NUREG-1537, Part 1, has been reproduced

from the best available copy.
CONTENTS

Page

ABSThRACT .......................... iii

ABBREVIATIONS ............ ;

INTRODUCTION ............ ; xv

Background . xv

Document Structure . .......................................... xviii

General Requirements ............. - ; oua

Contributors ... vi

References ............... xxT

I THEDFACUT ........... - ]-1

1.1 Introduction .......... -. - 1-1

1.2 Summary and Conclusions on Principal Safety Considerations ..-- 1-1

1.3 General Description ..................... .... 1-2

1.4 Shared Facilities and Equipment ..........................-. 1-2

1.5 Comparison With Similar Facilities. 1-3

1.6 Summary of Operations ................................. 1-3

1.7 Compliance With the Nuclear Waste Policy Act of 1982 .1-3

1.8 Facility Modifications and History .1-3

2 SITE CHARACTERISTICS ............... ........ 2-1

2.1 Geography and Demography ....................... 2-1

2.1.1 Site Location and Description .......... ............. 2-1

2.1.2 Population Distribution ........................ 2-2

2.2 Nearby Industrial, Transportation, and Military Facilities . . 2-2

2.2.1 Locations and Routes ... ..................... 2-3

2.2.2 Air Traffc ................................. ;;. 2-3

2.2.3 Analysis of Potential Accidents at Facilities . ; . '. 2-4

2.3 Meteorology ..................... . . 2-4

2.3.1 General and Local Climate ........ .......... 2-4

2.3.2 Site Meteorology ...................................... 2-4

2.4 Hydrology 2-5

2.5 Geology, Seismology, and Geotechnical Engineering . .2-5

2.5.1 Regional.Geology .......... 2-6

2.5.2 Site Geology ........................ . 2-6

2.5.3 Sesmicity . . .2-6

2.5.4 Maximum Earthquake Potential . . . 2-6

-2.5.5 Vibratory Ground Motion . . ............... 2-6

2.5.6 Surface Faulting . . .2-7

2.5.7 Liquefaction Potential . . .2-7

2.6 Bibliography .. 2-7

REV. 0,2196 v .2STANDARD FORMAT Am CONTENT
CONMIU

page

3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS ..... 3-1

3.1 Design Criteria ............................................ 3-1

3.2 Meteorological Damage .............. ....................... 3-2

3.3 Water Damage .............. .............................. 3-3

3.4 Seismic Damage ........................................... 3-3

3.5 Systems and Components .................................... 3-5

3.6 References ............................................... 3-5

4 REACTOR DESCRIPTION ............... .................... 4-1

4.1 Summary Description ...................................... 4-1

4.2 Reactor Core ............................................. 4-1

4.2.1 Reactor Fuel ............................................ 4-2

4.2.2 Control Rods .......................... 4-3

4.2.3 Neutron Moderator and Reflector ........ .................... 4-4

4.2.4 Neutron Startup Source ............... .................... 4-5

4.2.5 Core Support Structure ............. ....................... 4-5

4.3 Reactor Tank or Pool ...................................... 4-6

4.4 Biological Shield ......................................... 4-7

4.5 Nuclear Design . ......................................... 4-8

4.5.1 Normal Operating Conditions .......... ...................... 4-9

4.5.2 Reactor Core Physics Parameters ........... ................ 4-10

4.5.3 OperatingLimits .... ..................................... 4-11

4.6 Thermal-Hydrmlic Design ........... ...................... 4-12

Appendix 4.1 Regulatory Guide 2.1

S REACTOR COOLANT SYSTEMS ............................. 5-1

5.1 Summazy Description ................. ...................... 5-2

5.2 Primary Coolant System .............. ....................... 5-2

5.3 Secondary Coolant System ................................... 5-5

5.4 Prinmay Coolant Cleanup System .............................. 5-8

5.5 Primary Coolant Makeup Water System ....... .................. 5-9

5.6 Nitrogen-16 Control System .......... ....................... 5-10

5.7 Auxiliaiy Systems Using Primary Coolant ...... ................. 5-12

5.8 Reference ............................................... 5-13

6 ENGINEERED SAFETY FEATURES ........... ................ 6-1

6.1 Summary Description ...................................... 6-5

6.2 Detailed Descriptions ............... ...................... 6-5

6.2.1 Confinement ............................................ 6-6

6.2.2 Containment ............................................ 6-7

6.2.3 Emergency Core Cooling System ........ ..................... 6-9

6.3 References ............................................. 6-10

NUREG-1537,PARTI vi REV. 0. 2i96
COMM

Page

7 INSTRUMENTATION AND CONTROL SYSTEMS . ............... 7-1

7.1 Summary Description ...................................... 7-2

7.2 Design of Instrumentation and Control Systems ....... 7-3

7.2.1 Design Criteria ............................................ 7-3

7.22 Design-BasisRequirements ....................... - 7-4

7.2.3 System Description ;.. 7-6

7.2.4 System Performance Analysis ...................... 7-6

7.2.5 Conclusion .................... - 7-7

7.3 Reactor Control System .................... 7-7

7.4 Reactor Protection System . ............. 7-10

7.5 Engineered Safety Features Actuation Systems ............ ........ 7-13

7.6 Control Console and Display Instruments . ..................... 7-14

7.7 Radiation Monitoring Systems ................. ..... 7-15

7.8 Bibliography . ..................................... 7-16

Appendix 7.1 Regulatory Guide 1.152, Revision 1

Appendix 7.2 Amendment No. 29 to Facility Operating License No. R-3 8

Appendix 7.3 AmendmentNo. 30 to Facility Operating License No. R-2

Appendix 7.4 NRC Generic Letter 95-02

8 ELECTRICAL POWER SYSTEMS ............. ............... 8-1

8.1 Normal Electrical Power Systems ............................... 8-1

8.2 Emergency Electrical Power Systems ................................. 8-2

9 AUXILIARY SYSTEMS .. i-.9-1-. . 9 TCIIR SSEM .................................. ;- 9-1

9.1 Heating, Ventilation, and Air Conditioning Systems ...... .......... 9-3

9.2 Handling and Storage of Reactor Fuel ........................... 9-3

9.3 Fire Protection Systems and Programs ........................... 9-5

9.4 Communication Systems .................. 9-6

9.5 Possession and Use of Byproduct, Source, and Special

Nuclear Material ......... ; 9-6

9.6 Cover Gas Conirol in Closed Primary Coolant Systems ..... ......... 9-7

9.7 Other Amiliary Systems ....................... ;-. -. 9-8

9.8 References .. ; 9

Appendix 9.1 Regulatory Responsibilities for Byproduct Materials

in Non-Power Reactors -

Appendix 9.2 License Condition for Byproduct Material To Be

Irradiated in a Non-Power Reactor

-REv. 0. 2/% vii STMMU FOAT AM3 Coztir - REV. O. 2/96 -v,, - STAmARD FoRmATom Comnqr
I- - --

CONTNT

Page

10 EXPERIMENTAL FACITES AND UTILIZATION ............

10.1 Summary Description .....................................

10.2 ExperimentalFacilities ....................................

10.3 Experiment Review. ......................................

Appendix 10.1 Regulatory Guide 2.2

Appendix 10.2 Regulatory Guide 2.4

11 RADIATION PROTECTION PROGRAM AND

WASTE MANAGEMENT ..................................

10-1

10-3

10-5

104

11-1

11.1

11.1.1

11.1.2

11.1.3

11.1.4

11.1.5

11.1.6

11.1.7

11.2

11.2.1

11.2.2

11.2.3

11.3

Radiation Protection ................ ............. 11-2

Radiation Sources ............ ................. 11-2

Radiation Protection Program ............................. 11-6

ALARA Program. ............................. 1-8

Radiation Monitoring and Surveying ....... 11-8 1................. 

Radiation Exposure Control and Dosimetry ......... .......... 11-9

Contamination Control .............................. 11-11

Environmental Monitoring ............................. 11-12

Radioactive Waste Management ........................... 11-13

Radioactive Waste Management Program ......... .......... 11-13

Radioactive Waste Control ............................. 11-14

Release of Radioactive Waste ............................. 11-16

Bibliography ......... .................... 11-17

12 CONDUCT OF OPERATIONS ..

12.1 Organization ..............

12.1.1 Structure.................

12.1.2 Responsibility .............

12.1.3 Staffing ..................

12.1.4 Selection and Training of Persor

12.1.5 Radiation Safety ...........

12.2 Review and Audit Activities ...

12.2.1

12.2.2

12.2.3

12.2.4

12.3

12.4

12.5

12.6

12.7

12.8

12.9

12.10

Composition and Qualifications

Charter and Rules ..........

Review Function ...........

Audit Function ............

Procedures ................

Required Actions ...........

Reports ..................

Records ..................

Emergency Planning .........

Security Planning ...........

Quality Assurance .......... Operator Training and Requalifli

............................

............................

............................

............................

............................

ie ........................

............................

............................

...........................

............................

............................

............................

............................

............................

............................

............................

............................

............................

............................

cation ......................

12-1

12-1

12-1

12-2

12-2

12-2

12-2

12-3

12-3

12-4

12-4

12-4

12-4

12-5

12-5

12-6

12-6

12-6

12-7

12-7

NURBG-1537,PMT I vIu REv. 0,2/96

NUREG- 1 537, PART I wiil REV. O. 2/96
CONTENS

Page

12.11 StartupPlan ................... . 12-12

12.12 Environmental Reports .; .. 12-14

12.13 References ........... ;. ......... 12-15

Appendix 12.1 Regulatory Guide 2.6

Appendix 12.2 NUREG-0849

Appendix 12.3 Rerlato y Guide 5.59, Revision 1

Appendix 12.4 Regulatory Guide 2.5

13 ACCIDENTANALYSES '.. .. 13-1

13.1 Accident-Initiating Events and Scenarios ...... ............... 13-3

13.1.1 MaxinunHypotheticalAccident ............................. 13-3

13.1.2 Insertion of Excess Reactivity ............ .................... 13-4

13.1.3 Loss of Coolant ............. .......................... 13-4

13.1.4 Loss of Coolant Flow ................................ 13-5

13.1.5 Mishandling orMalfimction of Fuel... . - 13-5

13.1.6 Experiment Malfunction ................................ 13-5

13.1.7 Loss ofNormal Electrical Power ...... - 13-6

13.1.8 ExternalEvents ............. 13-6

13.1.9 Mishandling or Malfunction ofEquipment . . ..... 13-6

13.2 Accident Anaiysis and Deterinination of Consequences ..... ..... 13-7

13.3 Summary and Conclusions . -- 13-8

13.4 Bibliography ............ 13-9

14 TECHNICAL SPECIFICATIONS ...................

Reference .........................................

Appendix 14.1 Format and Content of Technical Specifications

14-1

14-2

15 FINANCIAL QUALIFICATIONS ...................

15.1 Financial Ability To Construct a Non-Power Reactor ....

15.2 Financial Ability To Operate aNon-Power Reactor ......

15.3 Financial Ability To Decommission the Facility .........

15.4 Reference .....................................

15-1

15-1

15-2

154

15-5

16 OTHER LICENSE CONSIDERATIONS .........

16.1 Prior Use ofReactor Components ............

16.2 Medical Use ofNon-PowerReactors ...........

Appendix 16.1 Amendment No. 27 to Facility Operating

License No. R-37

16-1

16-1

16-2

REv. 0,2196 ix STANDARD FOL4T ACONWNT

REV.0, 2196 - K STM;AmR FORMAT AND CoUNTr
CoNTENrs

page

17 DECOMMISSIONING AND POSSESSION-ONLY

LICENSE AMENDMENTS ................................. 17-1

17.1 Decommissioning ....................................... 17-1

17.1.1 PrelininayDecommissioningPlan ........................... 17-1

17.1.2 Submittal of the Decommissioning Plan . ................ 17-2

17.1.3 Decommissioning Alternatives ............................. 17-4

17.1.4 Release Criteria and Final Survey ............ ............... 17-5

17.1.5 Format and Content of Decommissioning Plan ....... .......... 17-6

17.2 Possession-Only License Amendment ....... ................. 17-6

17.2.1 An Application for a Possession-Only License Amendment .... ... 17-7

Appendix 17.1 Format and Content of Decommissioning Plan

for Non-Power Reactors

18 HIGHLY ENRICHED TO LOW-ENRICHED

URANIUM CONVERSIONS ......... ....................... 18-1

Appendix 18.1 Format and Content for HEU to LEU Conversions

at Non-Power Reactors

APPENDIX A APPLICABILITY OF SELECTED REGULATIONS IN

TITLE 10, CHAPTER I OF THE CODE OF FEDERAL

REGULA TIONS TO NON-POWER REACTORS

NUREG-1537, PART I x REV. 0. 2/96

NUREG-1 537, PART I x REV. O. 2/96
ABBREVIATIONS

ACRS Advisoty Committee on Reactor Safeguards

AEA Atomic Energy Act of 1954, as amended

AEC Atomic Energy Common

AGN AerojetGeneral Nucleonics

ALARA as low as is reasonably achievable

ANS American Nuclear Society

ANSI American National Standards Institute

ASLAB Atomic Safety and icensing Appeal Board

ATR advanced test reactor

BNCT boron neutron capture therapy

BSF bulk shielding ficility

CAM continuous air monitor(ng)

CFR Code of Federal Regulallons

CP construction permit

CP-5 Chicago Pile #5

CP-11 ChicagoPile#l1

DNBR departure from nucleate boiling ratio

DOE Department of Energy

DOT Department of Transportation

DP decommissioning plan

EA evrnetlassmn

EC effluent concentration

ECCS emergency core cooling system

ER environmental report

ESF engineered safety feature

EPRI Electric Power Research Institute'

EPZ emergency preparedness zone

FDA Food and Drug Administration

FLIP fuel lifetime improvement program

FR FederalRegiser

FSAR final safety analysis report

GA General Atomics

HEU highly enriched uraniunm

HVAC heating, ventilation, and air conditioning

. REV. 0,2/96 - 02STADARD FORMAT AND COmEwT
- -

ABBREVIATIONS

IAEA International Atomic Energy Agency

I&C instnumentation and control

EInstitute of Electrical and Electronic Engineers

INEL Idaho National Engineering Laboratory

LCO limiting condition for operation

LEU low-enriched uranium

LOCA loss-of-coolant accident

LOFA loss-of-flow accident

LSSS limiting safety system setting

LW.R light-water reactor

MHA a hypothetical accident

MIT Massachusetts Institute of Technology

MPC maximum permissible concentration

MTR materials testing reactor

NFPA National Fire Protection Association

NRC Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation

NUMARC Nuclear Management and Resources Council

ORNL Oak Ridge National Laboratory

ORRR Oak Ridge Research Reactor

OSHA Occupational Safety and Health Administration

OWR Omega West Reactor

PSAR prelixninary safety analysis report

PULSTAR pulsed training assembled reactor

QA quality assurance

RAM remote area monitor

RCS reactor control system

RG regulatory guide

RO reactor operator

RPS reactor protection system

SAR safety analysis report

SI Internation System of Units

SL safety limit

SNM special nuclear material

SNAP system for nuclear auxiliary power

SPERT special power excursion reactor test

SRO senior reactor operator

NUREG-1537,PART I xii REv 0,2/96

NUREG-1537, PART I )d REV 0, 296
ABBREVIATONS

TLD thermmoluminescence dosimeter

TRIGA training reactor, isotope [production], General Atomics

TS technical specification

USGS U.S. Geological Survey

UTM Universal Transverse Mercator

REv. 0.2/96 xiil STANDARD FoIuAAT CoNmr

REV. 0, 2/96 ... STANDm oDiAND o-m-
INRODUCTION

Background

This document describes acceptable format and content of the safety analysis

report (SAR) to be submitted to the U.S. Nuclear Regulatory Commission (NRC)

by an applicant or licensee of a non-power reactor for a new license, license

renewal, or license amendment. A companion document, NUREG-1537, Part 2

(Standard Review Plan), gives criteria to assist NRC staff reviewers in effecting

comparable, complete, and consistent reviews of licensing applications'for nonpower reactors. Applicants could peruse the Standard Review Plan to gain'firther

ingt into the review process for finding non-power reactor applications

acceptable.

NRC published several documents that give guidance that is applicable to

commercial power reactors. In 1972, to help commercial power applicants

prepare SARs for operating licenses for power reactors, NRC clarified the format

and content of SARs for light-water reactors (LWRs) by issuig Regulatory Guide

(RG) 1.70,'"Standard Format and Content of Safety Analysis Reports'for Nuclear

Power Plants (LWR Edition)," with'revisions in 1973,1975, and 1978. In 1975,

NRC issued NUREG-751087, "Standard ReiewPlan forth&Re~iew'of Safety

Analysis Reports for Nuclear Power Plants," to assist the NRC staff in effecting'

comparable, complete, and consistent reviews of SARs for nuclear power plants.

In 1982, the staff completely revised the earlier Standard Review Plan'(NUREG- -

75/087) and published the revision as NUREG-0800. In 1987, the staff revised

NUREG-0800. 7- -

These documents were developed specifically for LWR nuclear power plants.

Applicants who would use these documents to prepare SARs fbr non-power

reactor facilities and NRC staff who would use them to review these SARS may

find it very cumbersome because ofthe great differences in complexity and hazards

between non-power reactors and nuclear power plants. Thus, the NRC staff

started this program to document guidance applicable to non-power reactors. The

guidance herein is based on the Code of Federal Regulations, Thie 10, Section

50.34 (10 CFR 50.34) which describes the information to be supplied in a SAR.'

All reactors, both power and non-power, are licensed to operate as utflization

facilities under Title 10 in accordancewith the Atomic EneryAct of 1954, as

amended (AEA or Act). The AEA was written to promote the development and

use of atomic energy for peacefil purposes and to control and limit its radiological

hazards to the public. These purposes are expressed in paragraph 104 of the Act

for non-power reactors, which states that utilization facilities for research and

development should be regulated to the minimum extent consistent with protecting

the health and safety of the public and promoting the common defense and

REV. 0. 2/% xv STANDARD FORMAT AND CONIEwT
_ _-------

security. These concepts are promulgated in 10 CFR 50.40 and 50.41, and in

other parts of Title 10 that deal with non-power reactors. The licensed thermal

power levels of non-power reactors are several orders of magnitude lower than

current power reactors. Therefore, the accumulated inventory of radioactive

fission products in the fuel Cm core) of non-power reactors is proportionally less

than power reactors and requires less stringent and less prescriptive measures to

give equivalent protection to the health and safety of the public. Thus, even

though many of the regulations of Title 10 apply to both power and non-power

reactors, the regulations may be implemented in a different way for each category

of reactor and are intended to be consistent with protecting the health and safety of

the public. Because the potential hazards may also vary widely among non-power

reactors, regulations also may be implemented in a different way within the nonpower reactor category.

Sections 50.20 through 50.22 of Title 10 specify two classes of reactor licenses to

be issued to applicants by the NRC: Class 104 (medical therapy and research and

development facilities) and Class 103 (commercial and industrial facilities). These

classes derive from definitions in the AEA. Non-power reactors are designed and

operated for medical therapy, research, development, and education. Non-power

reactors consist of testing facilities (also called "test reactors" in some regulations)

which are defined in 10 CFR 50.2, and research reactors, which are defined in

10 CFR 170.3.

Currently, all non-power reactors are licensed as Class 104 facilities. However,

NRC recognizes that a non-power reactor for commercial purposes could be

licensed as a Class 103 facility, and thus, 10 CFR 50.22 contains criteria for

judging if a non-power reactor is a Class 103 facility.

A Class 104 non-power reactor can be licensed as a Class 104a facility for

conducting medical therapy or as a Class 104c facility for conducting research and

development. One non-power reactor is licensed as both a Class 104a and 104c

facility. All other non-power reactors are licensed as Class 104c facilities.

Most of the design, operation, and safety considerations for non-power reactors

apply to both test and research reactors. All non-power reactor applicants should

be guided by the format and content for licensing applications in this document.

Test reactors are subject to additional requirements, such as preparation of an

environmental impact statement, conduct of licensing hearings, and review by the

Advisory Committee on Reactor Safeguards (ACRS).

The issue of what standards to use in evaluating accidents at a research reactor

was discussed in an Atomic Safety and Licensing Appeal Board (ASLAB) decision

issued May 18, 1972, for the research reactor at Columbia University in New York

City. ASLAB stated that "as a general proposition, the Appeal Board does not

NURJEG-1537, PART 1 XVI REv. 0,2/96

NUREG- 1537, PART I xvi REV. O. 2r96
INTRoDUCIoN

consider it desirable to use the standards of 10 CFR Part 20 for evaluating the

effects of a postulated accident in a research reactor inasmuch as they are unduly

restrictive for that purpose. The Appeal Board strongly recommends that specific

standards for the evaluation of an accident situation in a research reactor be

formulated.' The staff has not found it necessary to conform to that

recommendation to develop separate criteria for the evaluation of research reactor

accidents, since most research reactor accidents evaluated to date have been within

the 10 CFRPart 20 criteria.

The pncipal safety issues that differentiate test reactors from research reactors are

the reactor site requirements and the doses to the public that could result from a

serious accident. For a research reactor, the results of the accident analysis have

generally been compared with 10 CFR Part 20 (10 CFR 20.1 through 20.602 and

Appendices for research reactors licensed before January 1, 1994, and 10 CFR

20.1001 through 20.2402 and Appendices for research reactors licensed on or

.after January 1, 1994). For research reactors licensed before January 1, 1994, the

doses that the staff has generally found acceptable for accident analysis results for

research reactors are less thanr 5 rem whole body and less than 30 rem thyroid for

occupational exposure, and less than 0.5 rem whole body and less than 3 rem

thyroid for members of the public. For research reactors licensed on or after,

January 1, 1994, occupational exposure is discussed in 10 CFR 20.1201 and public

exposure is discussed in 10 CFR 20.1301. In several instances, the staff has -

accepted very conservative accident analyses that exceed the 10 CFRPart 20 dose

limits discussed above. -

If the facility conforms to the definition of a test reactor, the doses should be

compared with 1O CFRPart 100. As discussed in the footnotes to 10 CFR

100.11, the doses given in 10 CFRPart 100 are reference values. Any further

references to 10 CFR Part 100 in this document apply to test reactors only.

The hazards from non-power reactors, compared with power reactors, range from

small to insignificant. After licensing almost 150 non-power reactors, the NRC

staff has developed guidelines and criteria for use in concluding that a f&lity,

function, or procedure provides reasonable assurance that the public will not

receive a radiation dose that exceeds regulatory limits.

The regulations in 10 CFR 2.105(c) for the initial licensing of a research reactor

facility do not preclude a joint application for a construction permit and the initial

operating license. If well planned, the frnml facility design and the final SAR

descriptions, analyses, and conclusions will not differ sigiificantly from those in the

initial application, and a one-step licensing procedure can be undertaken. To

initiate this process, the applicant should request both a construction permit and an

operating license to be issued when construction and operating readiness are

acceptable to NRC. The applicant should submit only one SAR that is complete,

REv 0,2/96 XVII STANDARD FORMAT AND CON1ENT

REV 0,21V6 XVII STANDARD FORMAT AND CoNTEr
appropriate, and acceptable for both permits. This will enable NRC to publish a

joint notice of intent in the Federal Register at the construction permit stage that

includes issuance of the operating license when appropriate. The joint application

and joint notice procedure streamlines the licensing process. If a final SAR

documenting changes during construction is submitted, it must demonstrate that

the facility design and the safety conclusions of the previous SAR documentation

are unchanged.

The design information in an SAR should reflect the current state of the facility

design, or the current as-built system at the time of the submittal. If certain

information noted herein is not yet available because the design has not progressed

sufficiently, the SAR should contain (1) the criteria and design bases used to

develop the required information, (2) the concepts and alternatives under

consideration, and (3) the schedule for completing the design and submitting the

missing information. The SAR for a new facility should describe the current

design of the facility in sufficient detail to enable the reviewers to determine

whether or not the facility can be constructed and operated in accordance with

applicable regulations.

The licensing process conforms to the legislative requirement for minim

regulation stated in Section 104 of the AEA. A license for facility operation

constitutes the legal agreement between the licensee and NRC, and both parties

must adhere to it rigorously. Quite often, because of applicant choice, the

licensing process leads to two or more facilities with the same type of fuel and the

same intrinsic safety limits being licensed for operation at maxmum power levels

differing by at least a factor of 10. The resultant difference between licensed

operating conditions and safety limits may vary by at least an order of magnitude.

However, each facility is obligated to adhere to its own license conditions.

Document Structure

Parts I and 2 of this document are complementary; titles and numbers of sections

correspond to the SAR sections.

The structure of the document is summarized below. The general requirements of

the safety analysis are presented along with information on the purpose,

applicability, and use of this document.

Chapter 1 summarizes the principal design bases and considerations, general

descriptions of the reactor facility that illustrate the anticipated operations, and the

design safety considerations, including the limiting potential accidents. This

chapter should summarize the detailed information found in subsequent chapters of

the SAR.

NLJRBG-1537, PART 1 XVIII REV 0,2/96 NLTREG- 1537, PART I XVIII REV 0, 2196
INMODtUarON

Chapter 2 describes the bases for the site selection and describes the applicable site

characteristics, including geography, demography, meteorology, hydrology,

geology,I seismology, and interaction with nearby installations and facilities.

Chapter 3 describes the design bases and facility structures, systems, and

components, and the responses to environmental factors on the reactor site (e.g.,

floods).

Chapter 4 describes the design bases and the functional characteristics of the

reactor core and its components. In this chapter, the safety considerations and

features of the reactor are discussed.

Chapter 5 lists the design bases and describes the fiinctions of the reactor coolant

and associated systems at the facility, including the primary and secondary systems

as applicable, and'coolant makeup and purification systems. The chapter also

describes provisions for adequate heat removal while the reactor is operating and

while it is shut down.

Chapter 6 lists the design bases and describes the functions of engineersafety

features (ESFs) that may be required to mitigate consequences of postulated,

accidents at the facility. This includes design-basis accidents and a maxim

hypothetical accident (MHA). The MHA, which assumes an incredible failure that

can lead to fuel cladding or to a fueled experiment containment breach, is used to

bound credible accidents in the accident analysis.

Chapter 7 lists the design bases and describes the functions of the instrmentation

and control systems and subsystems at the facility, placing emphasis on safetyrelated systems and safe reactor shutdown.

Chapter 8 lists the design bases and describes the functions of the normal and

emergency (if applicable) electrical power systems at the facility.'

Chapter 9 lists the design bases and describes the functions of such auxiliary

systems at the facility as heating, ventilation, air exhaust, air conditioning, service

water, compressed air, and fuel handling and storage.

Chapter 10 lists the design bases and aescribes the functions of experimental

facilities. Non-power reactors are designed with irradiation capabilities for

research, education, and technological development. This chapter discusses the

cha teristics of experiment and irradiitionifacilities on the basis of the proposed

experimental programs.

Chapter 11 lists the design bases and describes the functions of the radiation

protection and the radioactive waste management programs at the facility. This

REV 0,2196 

XIX

0, 219 Xlx -
chapter also describes the control of byproduct materials produced in the reactor

and utilized under the 10 CER Part 50 reactor operating license. The description

of the radiation protection program should include health physics procedures,

monitoring programs for personnel exposures and effluent releases, and assessment

and control of radiation doses, both to workers and the public. The program to

maintain radiation exposures and releases as low as is reasonably achievable

(ALARA) includes the control and disposal of radiological waste from reactor

operations and from experimental programs.

Chapter 12 lists the bases and describes the functions of plans and procedures for

the conduct of facility operations. These include discussions of the management

structure, personnel training and evaluation, provisions for safety review and

auditing of operations by the safety committees, and other required functions, such

as reporting, security planning, emergency planning, and planning for reactor

startup.

Chapter 13 lists the bases, scenarios, and analyses of accidents at the reactor

facility, and describes an MHA, which may include a fission product release, and

radiological consequences to the operational staff reactor users, the public, and

the environment. The function of ESFs is discussed in the accident analysis, as

applicable.

Chapter 14 presents the technical specifications, which state the operating limits

and conditions and other requirements for the facility to acceptably ensure

protection of the health and safety of the public.

Chapter 15 concerns financial qualifications of the non-power reactor applicant for

initial construction, continuing operations, and decommissioning.

Chapter 16 discusses other license considerations, such as prior reactor use and the

use of the reactor for medical therapy. Issues not discussed elsewhere in the SAR

are covered in this chapter.

Chapter 17 gives guidance on decommissioning. This includes the development of

a decommissioning plan and the preparation of an amendment request to amend a

license from operating to possession-only status.

Chapter 18, which discusses the conversion of the reactor from highly enriched

uranium (HEU) fuel to low-enriched uranium (LEU) fuel, includes topics covered

in Chapters I to 17 as related to HEU to LEU conversions.

Appendix A lists regulations in selected parts of Title 10 that apply to research and

test reactors.

NUREG-1537,PRT I REv 0.2/96
INIRODUCTON

General Requirements

Section 50.34 of rtIe 10 ofthe Code ofFederalRegukitionsrequires each

applicant for a license to submit an SAR in the application. An SAR is clearly

required for initial application for a license. Although no regulations apply

specifically to an SAR submitted for renewing a non-power reactor license, the

NRC staff can most effectively evaluate an application to renew a facility license

from an up-to-date SAR -

The SAR performs the following important functions:

* Gives a complete description of the facility.

i Documents the design bases of the facility.

i Demonstrates and documents that the facility is designed and can be

operated in a manner consistent with applicable regulations so that the

health and safety of the'public, the facility staff and users, and the

environment are protected.

* Documents the limits, restrictions, administrative controls, and planned

conduct of operations of the facility.

i Iicludes technical specifications based on the SAR. (The technical

specificationsi express an agreement between NRC and the applicant on

how the fiyw l be managed and operated to ensure the protection of

'the health and safety of facility personnel and the public, as well as

protection of the environment.)

The SAR contains the formal documentation for a facility, presenting basic

-information about'the design bases, and the considerations and reasoning used to

support the applicant's conclusion that the facility can be operated safely. The

descriptions and discussions therein also support the assumptions and methods of

analysis of potential 'accidents, including the MHA, and the design of any ESFs

used to mitigate accident consequences.

The SAR is the basic document that gives the NRC justification for licensing the

facility. It gives information for understanding the design bases for the 10 CFR

50.59 change process, for training reactor operators, for preparing reactor

operator licensing exminations, and for preparing for NRC inspections. For these

reasons, and for others, it is important that the SAR remain an accurate, current

description of the facility. Even'though regulations do not require the licensee for

a non-power reactor to periodically update the SAR (as is required in 10 CFR

50.71 (e) for power reactors), the NRC staff encourages non-power reactor

REV.~~ 0,29_TNADFRA N or~ . REV. 0, 2/96 STAmARD FoRmAT AND CoNTEm
licensees to maintain current SARs on file at NRC after initial licensing or license

renewal by submitting replacement pages along with applications for license

amendment and along with the annual report that summarizes changes made

without prior NRC approval under 10 CFR 50.59.

An applicant for license renewal should address all applicable topics in this

document by submitting an updated, complete SAR to account for any facility

changes and any new regulatory requirements. Updating the SAR as described

above will not completely eliminate the need to update and rewrite sections of the

SAR at license renewal; it can, however, reduce the amount of resources needed to

update the SAR. Licensees should review the license renewal requirements of

10 CFR 2.109 at least a year before the expiration date of a facility operating

license and should contact NRC for any additional guidance that may be needed.

Purpose of the Format and Content Guide

This guide will help the applicant ensure the completeness and uniformity of the

information submitted, assist the NRC staff and others in locating the information,

and aid in reducing review time.

Applicability of the Format and Content Guide

The NRC staff recommends this guide for license applications for new non-power

reactors and for license renewal applications for existing non-power reactors. This

document also gives guidance to licensees preparing SARs for other licensing

actions, such as license amendments. This guide should help licensees prepare

complete packages and, thus, should reduce potential delays caused by NRC

requests for additional information. Applications for license amendments should

be written in accordance with applicable sections of the guide; however, a

complete revision of the existing SAR should not be required in support of such an

application. For license amendment requests, the corresponding sections of the

SAR should be amended and submitted to NRC along with the amendment

application. A complete revision of the SAR is strongly encouraged for license

renewal. This format can be applied to all NRC-regulated non-power reactors.

However, license applicants for non-power reactors with power levels above

several tens of megawatts or with novel design features should contact the NRC

staff to determine if additional guidance is needed.

The NRC staff recognizes that not every suggestion given here will be applicable

to every non-power reactor. This problem is inherent to writing a single guidance

document for reactors ranging from an AGN design with 0. 1-watt thermal power

to the heavy-water tank test reactor at the National Institute of Standards and

Technology at 20 megawatts thermal power. As applicants consult this document,

they will identify guidance that they believe is not applicable to their particular

NUREG-1537, PART I REV. O. 2196
INTRODUCI1ON

reactor design. Applicants should carefully consider what guidance is applicable to

their reactor design. The applicant does not need to discuss the reasoning for not

providing all of the information suggested in this document. However, the

applicant should be able to justify such deletion upon request of the NRC reviewer.

Use of the Standard Format and Content Guide

Although applicants are not required to consult this guide in preparing the

application for a license or license amendment, the NRC staff strongly encourages

its use because all applications vill be reviewed and evaluated on the basis of their

technical content and completeness. Upon receiving an application, the NRC staff

will review and evaluate the SAR against the standard review plan (NUREG-1537,

Part 2) to determine if the SAR contains the information necessary to form the

bases for the staff findings required for the issuance or renewal of an operating

license or granting of a license amendment.

Physical Specifications of the Application

Style and Composiion

The applicant should

Clearly and concisely state the technical bases to support the adequacy of

designs or design methods.

* Include a contents page.

Include the topics and headings at least to the level of headings with three

digits (e.g., 2.4.2).

* NPlace an index of key items as back matter, if desired.

* Add appendices for supplemental information not explicitly discussed in

this guide. Such information could comprise summaries of the responses to

NRC regulatory guides or proposed regulations, and supplemental

information on calculational methods or design approaches.

* Avoid duplicating information. Similar information may be requested in

various parts because it is relevant to more than one portion of the facility

or analysis. However, this information should be presented in the principal

area of the document that deals with the topic and thereafter appropriately

referenced in the' other parts of the safety analyses. For example, where

piping and instrumentation diagrams for the same system are needed in

REV 0,2/96 .xx1n STANDARD FORMAT AND CorENT

REV 0, 2/96 X= STANDARD FORMAT AM CONTENT
M

more than one place in the document, the safety analyses may reference the

first location, provided all necessary information is presented.

* The number of significant figures in numerical values should reflect the

accuracy or precision to which the number is known. Where possible,

estimate limits of error or uncertainty.

* Include equations for technical detail in safety analyses. Equations should

use standard technical conventions and symbols. All symbols should be

defined

* Specify measurements in the units used in the design of the facility. If the

facility was designed in English units, the measurements should be given in

English units first, followed by the SI [International System of Units (or

metric)] numerical equivalent in parentheses. Drawings and diagrams in

English units need not be changed to add SI units. If the facility was

designed in SI units, only the SI units need be given.

* Use abbreviations in a consistent manner throughout the safety analyses

and in a manner consistent with generally accepted use. Any abbreviations,

symbols, or special terms unique to the facility or not in general use should

be defined the first time they are used in the safety analyses.

* Submit three signed and notarized copies of the application in accordance

with 10 CFR 50.30, 10 CFR 50.4, and Generic Letter 84-18. Part 170 of

Title 10 of the Code of Federal Regulations presents information on

licensing and amendment fees.

Graphic Presentations

The applicant should use graphic presentations, such as drawings, maps, diagrams,

sketches, and tables to convey information more clearly or conveniently. All

information should be clearly legible and reproducible and all symbols should be

defined. Locate graphic presentations where the information they contain is

primarily discussed.

References

The applicant should list documents referenced under a heading or topic at the end

of the chapter or place them as footnotes on the page on which they are discussed.

If the former, cite references in the text parenthetically by author and date or by

reference numbers. If proprietary documents are referenced, cite a non-proprietary

summary of the document if available (see 10 CFR 2 790).

NUREG-1 537, PART I ,cV REv. 0,2/96
INTRODUMrON

Printing Specificaiions

The applicant should be guided by the following recommendations:

* Paper size should be 8-l2x l inches fortext pages and for most

drawings and graphics. If a larger page is needed, the finished copy should

not exceed 8-1/2 x 11 inches when folded.

* Maintain a margin of no less than 1 inch on the top, bottom, and binding

side of all pages submitted.

* Number pages with the digits corresponding to the chapter number

followed by a hyphen and a sequential number (e.g., the third page of the

discussion under Chapter 4, 'Reactor," should be numbered 4-3).

* Use paper and ink suitable in composition, paper color,-and ink density for

microflhuing or photocopying.

* Text pages may be single or double spaced using a suitable type face and

style for microfilming or photocopying.

* The document may be mechanically or photographically reproduced. All

pages of text may be printed on both sides of the paper. However, each

major section (contents, chapter, appendix, etc.) should start on a righthand page.

* Pages should be punched for standard three-hole loose-eaf binders.

Proceires for Uialing or RisingPages

The applicant should update or revise this guide by replacing pages. It is

recommended'that the applicant highlight the changed portion on each page by

placing a change indicator mark consisting of a bold vertical line drawn in the

margin opposite the binding margin (ioe., the outside margin). The line shoiild be

the same length as the portion changed. Replacement pages may be added in

response to NRC staff requests for additional information.

All changed pages should show the date of change and a revision or change

number. A guide with instructions for inserting, exchanging,' or removing pages

should accompany the'changed pages. If affected, revised contents pages should

be submitted.

REV. 0,2/96 .,ocv STmwFoiutT A?CONTENT

REV.0, 2/96 .Xocv STqmmD FoRmkT AND Com'ENr
- ~- W -

Other Forms of Presentation

Other forms of presentation may be used. However, under 10 CFR 50.4(c), the

applicant should contact the Information and Records Management Branch, Office

of Information Resources Management, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, to obtain specifications and copy requirements

before submitting materials other than paper.

Revisions

The format and content guide should be revised and updated periodically as

needed to clarify the content, to correct errors, and to incorporate modifications.

The revision number and publication date should be printed at the bottom of each

revised page. The revision numbers and dates need not be the same for all sections

because individual sections will be replaced with a newly revised section only as

needed. A list of affected pages will indicate the revision numbers for the current

sections. As necessary, the staff will make corresponding changes to the standard

review plan using these methods.

Contributors

This document was prepared by A. Adams, Jr., Senior Project Manager, NonPower Reactors and Decommissioning Project Directorate, Division of Project

Support, Office of Nuclear Reactor Regulation, U.S Nuclear Regulatory

Commission Major contributors to the document include the project manager,

S. Weiss, and M. Mendonca and T. Michaels, also of NRC; S. Bryan,

W. Carpenter, R Carter, D. Ebert, R. Garner, P. Napper, and P. Wheatley of the

Idaho National Engineering Laboratory (INEL) under contract to NRC; and

J. Hyder, J. Teel, and C. Thomas, Jr., of Los Alamos National Laboratory under

contract to INEL. Comments and suggestions for improving this document should

be sent to the Director, Non-Power Reactors and Decommissioning Project

Directorate, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001. Notices of errors or omissions should

be sent to the same address.

References

U.S. Nuclear Regulatory Commission, "Standard Format and Content of Safety

Analysis Reports for Nuclear Power Plants (LWR Edition)," 1972.

U.S. Nuclear Regulatory Commission, Generic Letter 84-18, "Filing of

Applications for Licenses and Amendments," July 6, 1984.

NUREG-1537, PART I REV. 0,2/96

PART I REV. O. 2196
INTRODUcrON

U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for

the Review of Safety Analysis Reports for Nuclear Power Plants (LWR Edition),"

1987.

U.S. Nuclear Regulatory Commission, NUREG-1 537, 'Guidelines for Preparing

and Reviewing Applications for the Licensing of Non-Power Reactors: Standard

Review Plan and Acceptance Criteria," Part 2, February 1996.

_~v 0_29 oiSADR o..~ N OTN

REV. O. 2t96 Anna STANDARD FmmAT AND CoNTENT
1 TIE FACILITY

In this chapter of the safety analysis report (SAR), the applicant should present an

introduction to the SAR and the facility. The introduction should state the

purpose of the SAR and briefly describe the application. Chapter 1 should contain

the following topics:

* introduction

* summary and conclusions of principal safety considerations

* general description of the facility,

* shared facilities and equipment

* - comparison with similar facilities

* -summary of operations

* compliance with the Nuclear Waste Policy Act of 1982

* facility modifications and history

1.1 Introduction

The applicant should state its name and description (e.g., university, government

agency, research institute, or company name) and should briefly state the purpose

and intended use of the facility, the geographical location of the facility, the reactor

type and power level, including principal inherent or passive safety features, and

any unique design features.' These topic's 'shoiuld be covered in full and referenced

to later chapters of the SAR.

1.2 Summary and Conclusions on Principal Safety

Considerations

The applicant should state safety criteria, the principal safety considerations, and

the resulting conclusions, including brief discussions of the following:

* consequences from'the operation and use of the non-power reactor, and

the methods used to ensure the safety of the reactor

safety ' considerations that influenced the selection of the facility site, the

type of reactor and fuel, the reactor thermal power level, the type of

building housing the reactor, and any special factors

* any inherent or passive safety features designed to contribute to facility

safety, protection of the health and safety of the public and staff; and

protection of the environment

* design features and design bases for any systems and components that

promote safe operation and shutdown of the facility

REV. 0. 2/96 - 1-1 STANDARD FORMAT AND CotrrErr

REV. O. 2/96- 1-1 . STANDmw FORMAT AND CoLTEN
CHAR I

potential accidents at the facility, including the maximum hypothetical

accident, and any design features that prevent accidents or mitigate the

potential consequences

These discussions need only be a general overview, with reference to the chapters

in which detailed analyses appear.

1.3 General Description of the Facility

The applicant should briefly describe the reactor facility as follows:

(1) geographical location; (2) principal characteristics of the site; (3) principal

design criteria, operating characteristics, and safety systems; (4) any engineered

safety features; (5) instrumentation, control, and electrical systems; (6) reactor

coolant and other auxiliary systems; (7) radioactive waste management provisions

(or system) and radiation protection; and (8) experimental facilities and

capabilities. The general arrangement of major structures and equipment should be

indicated with plan and elevation drawings. Safety features of the facility that are

likely to be of special interest should be briefly identified. Such items as unusual

site characteristics, the containment building, novel designs of the reactor, or

unique experimental facilities should be highlighted. The information and

discussions in this section in no way should substitute for the complete discussion

and analysis found in (and referenced to) subsequent chapters of the SAK

1.4 Shared Facilities and Equipment

The applicant should briefly describe the following:

* Systems and equipment that are shared with facilities not covered by the

SAR or the operating license. Examples of shared facilities and equipment

could be water purification systems; electrical supplies; heating, ventilation,

and air conditioning systems; and the building that houses the reactor

room.

* Any other reactor, subcritical assembly, irradiation facilities, or hot cell

located within the confinement or containment structures, or the restricted

area to which this SAR applies.

* Any safety barriers and any special isolation provisions for the shared

facilities and equipment.

Complete descriptions and any safety implications that result from sharing facilities

or systems should be evaluated in and referenced to the appropriate chapter of the

SAR

NUREG-I537,PMT I 1-2 REv 0,2196

NUREG- 1537, PART I 1-2 REV 0, 2196
THEFAarTy

1.5 Comparison With Similar Facilities

The applicant should describe briefy the principal similarities to other facilities,

particularly those either licensed by the U.S. Nuclear Regulatory Commission

(NRC) or designed and operated by the U.S. Department of Energy (DOE).

Comparisons should be made of the principal design parameters, reactor safety

systems, engineered safety features, and instrumentation and control systems. The

operating history of these facilities should be referenced briefly to demonstrate the

safety and reliability of the design. Design features, operations experience, and

tests and experiments from similar facilities could be referenced and used to

support analyses in appropriate chapters of the SAR.

1.6 Summary of Operations

The applicant should briefly discuss reactor operations, experimental programs,

and the mission of the reactor. The actual or proposed operations are important

for estimating parameters such as total operating time, power level, pulsed or

steady-state operation, and the amount and type of radioactive byproduct materials

produced. If the facility licensee is applying for license renewal, this section should

reflect current and proposed operational plans. If safety considerations analyzed in

later chapters of the SAR limit the operating schedule of the reactor, that fiat

should be noted here.

1.7 Compliance With the Nuclear Waste Policy Act of

1982

The applicant should briefly discuss how it meets the requirements of

Section 302(b)(1)(B) of the Nuclear Waste Policy Act of 1982 for disposal of

high-level radioactive wastes and spent nuclear fuel. This discussion should

include the contract arranged with DOE for return of the material. A copy of the

cover letter for the contract between the applicant and DOE should be included in

an appendix to the SAR

1.8 Facility Modifications and History

This section of the SAR applies primarily to existing facilities that are applying for

license renewal. This section has limited applicability to an application for initial

construction permit and operating license. The applicant should present a brief

history of the facility, including the dates of significant events, issuance of the

construction permit and operating license, and initial criticality. If the licensee has

experience with other research reactors, a brief description of this-inforrnation

should be presented. The SAR should indicate if the facility has not undergone

significant or safety-related physical or operational modifications since it was

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REV 0, 2/96 1-3 . STANDARD FoRmAT AND CoNTENT
- x -

CHAWPER I

initially licensed, or since the last renewal was issued. The SAR should reflect any

significant modifications made to the non-power reactor, programs, or schedules.

The modifications should be discussed briefly in this section in chronological order,

including the number and date of the license amendment. Changes performed

under the provisions of Section 50.59 of lithle 10 of the Code of Federal

Regulations (10 CFR 50.59) that affect the SAR descriptions should be provided.

If applicable, technical specifications changes should also be given.

NUREG-1537, PART I 1-4 REV. O. V96
2 SITE CHARACTERISTICS

In this chapter of the SAR, the applicant should discuss and describe the

geographical, geological, seismological, hydrological, and meteorological

haracteistics of the site and vicinity in -conjunction with present and projected

population distributions, industrial facilities and land use, and site activities and

controls.- In this presentation, the applicant should define the site characteristics

for use in design and analyses discussions in other chapters of the SAR, e.g.,

Chapter 3, 'Design of Structures, Systems and Components"; Chapter 11,

"Radiation Protection Program and Waste Management"; and Chapter 13,

"Accident Analyses." In 10 CFR 100.10, the staff gives factors to consider in

selecting sites for test reactors and related reactor design.

2.1 Geography and Demography

2.1.1 Site Location and Description

2.1.1.1 Specification and Location

The reactor should be located by latitude and longitude to the nearest second and

by Universal Transverse Mercator Coordinates [Zone Number, Northing, and

ating~ as found on U.S. Geological Survey (USGS) topographical maps) to the

nearest 100 meters. The State and county or other political subdivision in which

the site is located should be identified, including the location on a campus, if

applicable, as well as the location of the site with respect to prominent naural and

manmade features such as highways, rivers, lakes, reservoirs, and mountains.

2.1.1.2 Boundary and Zone Area Maps

Maps of the site area of suitable scale (with explanatory text as necessary) should

be included to show the boundaries and zones associated with the facility.

Boundaries and zones are defined in American National Standards

Institute/American Nuclear Society (ANSVANS) 15.7 (1977), "Research Reactor

Site Evaluation," Section 2, and in the 'documents in the bibliography related to

emergency preparedness at non-power reactors (NRC Regulatory Guide 2.6, NRC

NUREG-0849, ANSVANS 15.16). The'maps should clearly show the following:

the general area in which the reactor will be located and sufficient

secondary detailed maps to show the location of the reactor facility and

adjacent surroundings'

e location of the area directly under the NRC facility operating license

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CHAPT 2

0 location of the operations boundary

* location of the site boundary

* location of the rural zone up to a distance of 8 kilometers from the reactor

* location of the urban zone up to a distance of 8 kilometers from the reactor

* location of emergency preparedness zones (EPZs), as applicable

* true north

* highways, railways, and waterways that traverse or are in close proximity

to the site

* the general topography of the area near the reactor that could affect

diffusion and dispersion of airborne effluents, including buildings at least as

tall as the reactor building and any stacks or other air-exhaust facilities

2.1.2 Population Distribution

Population data presented should be based on the most recently available (last

decade or later) census data. Information on population distributions should be in

suitable form to use in dose analyses in Chapters 11 and 13, in which

potential doses down to a small percentage of 10 CFR Parts 20 or 100 may be

applicable.

On a map of suitable scale that identifies places of significant population grouping

(such as cities and towns) within an 8-kilometer radius, concentric circles should

be drawn, with the reactor at the center point, at distances of 1, 2, 4, 6, and 8

kilometers. The population in each area at the time of application and a projection

of the population in five years and at the end of the license period should be given.

The basis for population projections should be described. Information should be

given about the direction and distance of the nearest permanent residence to the

reactor and any reactor effluent exhaust points. Any part-time, transient, or

seasonal occupation of buildings should be described, such as classrooms or

dormitories on a university campus, giving best estimates of occupation times and

numbers of occupants.

2.2 Nearby Industrial, Transportation, and Military

Facilities

In this section, the applicant should establish whether the effects of potential

accidents in the vicinity of the reactor from present and projected industrial,

NUREG-IS37,PRT I 2-2 REV.O,2/96 NUREG- 1 537, PART I 2-2 REV. 0, V96
SITE CHARACrERISncs

transportation, and military installations and operations should be used in the safety

analyses and should establish the reactor facility design parameters related to

accidents selected. The applicant should consider all facilities and activities within

8 kilometers of the reactor. Facilities and activities at greater distances should be

included as appropriate to their significance of accident impact on the facility.

2.2.1 Locations and Routes

The applicant should submit maps showing the location and distance from the

reactor of all significant manufacturing plants; chemical plants; refineries; storage

facilities; mining and quarrying operations; military bases,' missile sites;

transportation routes (air, land, and water); transportation facilities (docks,

subways, highways, railways, and rail yards, anchorages, airports);-oil and gas

pipelines, drilling operations, and wells; and underground facilities used for such

purposes as fuel storage and storm-water runoff. These maps should'show any

other facilities that, because of the products manufactured, stored, or transported

there, may require consideration with respect to possible adverse effects on the

reactor. Any military firing or bombing ranges and any nearby aircraft flight,

holding, and landing patterns should be indicated on the maps.

The maps -should be clearly legible and of suitable scale to enable easy location of

the facilities and routes in relation to the reactor. All symbols and notations used

to depict the location of the facilities and'routes should be identified in legends or

tables.- Topographic features should appear on the maps in sufficient detail to .

illustrate the information presented and to support analyses of potential impacts on

the reactor facility.

2.2.2 'Air Traffic

Factors such as frequency and type of aircraft movement, flight patterns, local

meteorology, and topography should be considered for the following sites:

sites located within 8 kilometers of an existing or projected commercial or

military airport

* sites located between 8 and 16 kilometers from an existing or projected

commercial or military airport with more than approximately 200 d'

(where d is the distance in kilometers from the airport to the reactor site)

commercial or military aircraft movements per year.

Special consideration should be given when siting the facility within the trajectory

of a runway of any airport. The aiialysis should demonstrate that there is a low

potential that any aircraft, including general aviation aircraft, could affect the

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reactor or that the consequences from any aircraft-associated accident are already

bounded or considered in the accident analysis.

2.2.3 Analysis of Potential Accidents at Facilities

For each facility identified in Section 2.2.1, the applicant should provide an

analysis of possible effects on the reactor for postulated accidents or other events

that could occur at the facility. If a facility cannot affect the reactor, the applicant

should make a statement to that effect and give a basis for this statement.

2.3 Meteorology

In this section, the applicant should describe the meteorology of the site and its

surrounding areas. Sufficient data on average and extreme conditions should be

included to permit an independent evaluation by the reviewer.

23.1 General and Local Climate

The general climate of the region should be described with respect to types of air

masses, synoptic features (high- and low-pressure systems and frontal systems),

general and prevailing air-flow patterns (wind direction and speed), temperature

and humidity, precipitation (rain, snow, and sleet), and relationships between

synoptic-scale atmospheric processes and local (site) meteorological conditions.

References should indicate the climatic atlases and regional climatic summaries

used.

Historical seasonal and annual frequencies of severe weather phenomena, including

hurricanes, tornadoes, waterspouts, thunderstorms, lightning, and hail, should be

stated. The applicant should give the known and maximum annual frequency of

occurrence and time duration of freezing rain (ice storms) and dust (sand) storms

where applicable. The applicant should estimate the 100-year return wind speed.

The applicant should also estimate the weight of the 100-year return period

snowpack and the weight of the 48-hour probable maximum precipitation for the

site vicinity, if applicable, as specified by the USGS. Using these estimates for

Chapter 3, the applicant should calculate the design loads on the roof of the

reactor building, and compare them with local building codes for similar types of

structures.

2.3.2 Site Meteorology

In addition to discussing potential meteorological effects on the reactor facility, the

applicant should give sufficient information to support the dispersion analyses of

airborne releases from the facility. The applicant may need to evaluate potential

radiological effects in both the restricted and unrestricted areas in the reactor

NUREG-I537,PMT I 2-4 REV.O,2/96 NUREG- 1537. PART I 24 REV. O. 2196
SnE CHAxkCERSTcs

vicinity from routine releases during normal operations and from postulated

releases resulting from accidents. The analyses of potential doses from normal and

accident releases should be placed in Chapters'l I and 13, respectively. The

meteorological information used for both long-term and short-term dispersion

calculations, along with a description of the technical bases of the dispersion model

should be summarized. The continuing onsite measurements program or an

alternative source of meteorological information (e.g., National Weather Service

station) should be described; and plans for access to meteorological information

during the license period should be described. Description of the meteorological

program should include measurements made, locations and elevations of

measurements, description of instruments and their performance specifications, and

calibrations, type of data output, and data analysis procedures.-

2.4 Hydrology

In this section, the applicant should give sufficient information to allow an

independent hydrologic engineering review to be made of all hydrologically related

design bases, performance requirements, and bases for operation of structures,

systems, and components important to safety.

Sufficient information should also be given about the water table, groundwater,

and surface water features at the reactor site to support analyses and evaluations in

Chapters 11 and 13 of consequences of uncontrolled release of radioactive'material

from pool leakage or failure, neutron activation of soils in the vicinity of the

reactor, or deposition and migration of airborne radioactive material released to

the unrestricted area.

The effect of potential floods on sites along streams, rivers, and lakes should be

analyzed. Effects and consequences of a probable maximum flood, seiche, surge,

standing water, drainage'or seismically induced flood (such as might be caused by

dam failure) should be considered. Hazards of tsunami, river blockage, diversion

in the river system, or distant or locally generated "sea waves', should be described

to establish the suitability of a'site. The detail and extent of the considerations

should be commensurate with the potential consequences to the reactor and to the

public, 'the environment, and the facility staff.

2.5 Geology, Seismology, and Geotechnical Engineering

In this section, the applicant should detail the seismic and geologic characteristics

of the site and the region surrounding the site. The degree of detail and extent of

the considerations should be commensurate with the potential consequences of

seismological disturbance, both to the reactor facility and to the public from

radioactive releases.

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CHAPTER 2

2.5.1 Regional Geology

The applicant should discuss all geologic and seismic hazards within the region

that could affect the facility, and relate them to the regional physiography, tectonic

structures and tectonic provinces, geomorphology, stratigraphy, lithology, and

geologic and structural history and geochronology.

2.5.2 Site Geology

The applicant should discuss in detail the structural geology at the facility site,

including the relationship of site structure to regional tectonics, and should pay

particular attention to specific structural units of significance to the site such as

folds, faults, synclines, anticlines, domes, and basins. The applicant should also

discuss the geologic history of the site and should relate it to the geologic history

of the region.

2.5.3 Seismicity

The applicant should list all historically reported earthquakes that could have

reasonably affected the region surrounding the site. The list should include all

earthquakes of modified Mercalli intensity greater than IV or magnitude (Richter)

greater than 3.0 that have been reported in all tectonic provinces, any part of which

is within 200 kilometers of the site.

2.5.4 Maximum Earthquake Potential

The applicant should note the largest historic earthquake associated with each

geologic structure or tectonic province. If the earthquakes are associated with a

geologic structure, the applicant should evaluate the largest earthquake that could

occur on that structure on the basis of such considerations as the nature of

faulting, fault length, fault displacement, and earthquake history. If the

earthquakes are associated with a tectonic province, the applicant should identify

the largest historical earthquakes within the province and, whenever reasonable,

should estimate the return period for the earthquakes. Also, isoseismal maps for

the earthquakes should be presented.

2.5.5 Vibratory Ground Motion

The applicant should proceed from discussions of the regional seismicity, geologic

structures, and tectonic activity to a determination of the relation between

seismicity and geologic structures. The earthquake-generating potential of

tectonic provinces and any active structures should be identified. Finally, the

applicant should assess the ground motion at the site from the maximum potential

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NUREG- 1537, PART I 2-6 REV 0, 2/96
SITE CHAJXACIERISTICS

earthquakes associated with each tectonic province or geologic structure and

should consider any site-amplification effects. Using the results, the applicant

should establish the vibratory ground motion design spectrum.

2.5.6 Surface Faulting

The applicant should discuss any potential for surface faulting at the site, and

should list all historically reported earthquakes that can be reasonably associated

with faults, any part of which is within 8 kilometers of the site.

2.5.7 Liquefaction Potential

The applicant should discuss soil structure. If the foundation materials at the site

adjacent to and under safet-related structures are saturated soils or soils that have

a potential for becoming saturated, the applicant should prepare an appropriate

state-of-the-art analysis of the potential for liquefaction at the site. The applicant

should also determine the method of analysis on the basis of actual site conditions,

the properties of the reactor facilities, and the earthquake and seismic design

requirement for the protection of the public.

2.6 Bibliography

American National Standards Institute/American Nuclear Society, ANSIIANS

15.7, 'Research Reactor Site Evaluation," 1977.

American National Standards Institute/American Nuclear Society, ANSVANS

15.16, 'Emergency Planning for Research Reactors," 1982.

International Atomic Energy Agency, IAEA-TECDQC-348, 'Earthquake

Resistant Design of Nuclear Facilities With Limited Radioactive Inventory," 1985.

International Atomic Energy Agency, IAEA-TECDOC-403, "Siting of Research

Reactors," 1987.

U.S. Nuclear Regulatory Comnmission', NUREG-0849, 'Standard Review Plan for

the Review and Evaluation of Emergency Plans for Research and Test Reactors,"

1983.

U.S. Nuclear Regulatory Commission, NUREG/CR-2260, 'Technical Basis for

R.G. 1.145 Atmospheric Dispersion Models," 1981.

U.S. Nuclear Regulatory Commnission, Regulatory Guide 1.145, Rev. 1,

"Atmospheric Dispersion Models for Potential Accident Consequence Assessments

at Nuclear Power Plants," 1982.

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U.S. Nuclear Regulatory Commnission, Regulatory Guide 2.6, 'Emergency

Planning for Research and Test Reactors," 1983.

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NUTREG- I 537, PART I 2-8 REV 0, 2/96
3 DESIGN OF STRUCTURES, SYSTEMS, AND

COMPONENTS '

In this chapter of the SAR, the applicant should identify and describe the principal

architectural and engineering design criteria for the structures, systems, and

components that are 'required to ensure reactor facility safety and protection of the

public. The material presented should emphasize the safety and protective

functions and related design features that help provide defense in depth against

uncontrolled release of radioactive material. The bases for the design criteria for

some of the systems discussed in this chapter may be developed in 'other chapters

and should be appropriately cross referenced. For example, confinement or

containment design criteria may be summarized here and discussed in detail in

Chapter 6, "Engineered Safety Features."

3.1 Design Criteria

In this section the applicant should identify the structures, systems, and

components; modes of operation; location; type(s) of actuation; relative

importance in the control of radioactive material and radiation; applicable design

criteria; and chapter and section in the SAR where these criteria are applied in the

design of specific structures, systems, and components.

The design criteria should include applicable standards,' guides, and codes, for

example, American National Standards Institute/American Nuclear Society

(ANSIIANS) standards (see references); NRC regulatory guides (see Division 2

regulatory guides and 'Other Regulatory Guides of Possible Interest to Division 2

Recipients," which is a list attached to the Division 2 regulatory guides table of

contents); and national, State, and local building, plumbing, and electrical codes.

In this section the applicant should specif the design' criteria for the facility

structures, systems, and components. The description of the actual design should

be in the section or chapter that corresponds to the specificstructure, system, or

component. The design criteria should be both specific and general. The amount

of detail given should be related to the safety function of the structure, system, or

component. For example, general design criteria should include the following-.-

* Design for the complete range of normal expected reactor operating

conditions (e.g.; reactor power levels from cold subcritical conditions to

maximum allowed power level, associated radiation and system

temperature conditions, and allowed fuel storage and reactor operating

.configurations)..

* Design to cope with anticipated transients and potential accidents,'

including those discussed in Chapter 13, "Accident Analyses," of the SAR.

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CHAPTER 3

Anticipated transients and potential accidents should include malfunction of

any control function or other malfunction of a structure, system, or

component, experiment malfunction, single operator error, testing and

surveillance activity, reactor startup and shutdown, and power-level

change. These design criteria should be based on a systematic examination

of the most limiting transients and accidents to identify the needed facility

structures, systems, and components. They should ensure that each needed

structure, system, or component stays within acceptable operational and

safety limits for conservative assumptions of initial conditions, operating

history of the facility for the proposed license term, and required operating

characteristics. The most limiting conditions of each type should be

analyzed in detail in Chapter 13 of the SAR.

* Design redundancy for reactor protective and safety features, so that any

single failure of any active component will not prevent safe reactor

shutdown or result in unsafe conditions as verified by Chapter 13 analyses.

* Design to facilitate inspection, testing, and maintenance of the structures,

systems, and components whose integrity and reliability are important to

safe reactor shutdown and to the protection of the public, reactor facility

personnel, and environment.

* Provisions to avoid or mitigate the consequences of fires, explosions, and

other potential manmade or natural conditions.

* Quality standards commensurate with the safety function and the potential

risks. For example, fuel fabrication may be consistent with the applicable

guidance in ANSI/ANS 15.2-1990.

* Analyses and designs for meteorological, hydrological, and seismic effects.

* The bases for technical specifications necessary to ensure the availability

and operability of required structures, systems, and components.

3.2 Meteorological Damage

In this section the applicant should describe the design for the protection from

meteorological conditions of facility structures (e.g., buildings and cooling

towers), systems (e.g., ventilation systems), and components that are assumed to

be operable in the SAR. The design criteria should be based on data given in

Chapter 2, "Site Characteristics," on such factors as historical data on maximum

wind velocity, vertical velocity profiles, gust factors, applied loads, recurrence

intervals, tornado loadings, and snow and ice loads. The applicant may refer to

local building codes, standards, or other criteria to ensure that significant

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DESIGN OF STRUCrURES. SYSTEM, AND COMPONEB

meteorological damage to the facility is very unlikely. Further, the design criteria

should provide reasonable assurance that potential meteorological damage would

not significantly affect designed structures, systems, and components (i.e., they

would continue to perform necessary operational and safety functions). An

example would be consideration of adverse wind conditions that affect ventilation

systems. -The bases for appropriate technical specification surveillances to verify

capability and reliability of the design features should be given.

3.3 Water Damage

In this section the applicant should specifically describe the proposed site'and

facility designs to protect against water damage of the. structures, systems, and

components assumed to function in the SAR. This should include (I) the impact

on structures resulting from the force or submergence of flocding, (2) the impact

on systems resulting from instrumentation and control electrical ormeclanical

malfunction due to water, and (3) the impact on equipment, such as fans, motors,

and valves, resulting from degradation of the electromechanical function due to

water. This section should be based on historical data on the site with regard to

potential flooding and other hydrological conditions discussed in detail in Chapter

2. Design criteria for structures and systems that are based on information on

precipitation rates, ground water, accumulation of standing water, and drainage

rates should be included. The impacts of watersheds, flood plains, drainage

easements, and fluid supplies or conduits on reactor'operation, sife shutdown, and

control of radioactive material should also be included. Maps and other'

information in Chapter 2 to describe these features and characteristics around the

facility should be used.

The applicant should use local building codes or other applicable standards to'

ensure that significant water damage to the ficility'is very unlikely. Facility design

features (e.g., elevations, sumps, pumps, watertight doors, berms, and drains) may

be used to avoid or mitigate water damage to structures, systems, and components

important to safety. The applicant should show that the design features are

sufficient to avoid significant water damage to the facility during the projected

reactor license term.

The bases for any technical specifications required to ensure operability of

structures, systems, and components that ensure safe reactor shutdown should be

given.

3.4 Seismic Damage:

In this section the applicant should specify and describe the structures, systems,

and components that are required to maintain the necessary safety function if a

seismic event should occur, as well as the required facility seismic design criteria.

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CHAPTER 3

The seismic characteristics of the site should be summarized in Chapter 2. Seismic

design for non-power reactors should, at a minimum, be consistent with local

building codes and other applicable standards.

The reactor facility seismic design should provide reasonable assurance that the

reactor could be shut down and maintained in a safe condition or that the

consequences of accidents would be within the acceptable limits. For most NRC-

'licensed non-power reactors, this may involve analysis to show that the conditions

of the SAR for safe reactor shutdown remain valid for the potential seismic events

(e.g., reactor fuel fission product barrier and shutdown capability would not be

impaired). The applicant can also show that the radiological consequences of a

potential seismic event are bounded by the accident analyses in Chapter 13.

Such criteria for acceptable seismic performance have been established in

ANSI/ANS 15.7. With regard to seismic design, Section 3.2(2) of

ANSI/ANS 15.7 states, "(R)eactor safety related structures and systems shall be

seismically designed such that any seismic event cannot cause an accident which

will lead to dose commitments in excess of those specified in 3.1." For NRClicensed non-power reactors, 'any seismic event" should be the maximum historical

intensity earthquake in accordance with the guidance on the design-basis

earthquake in Section 3.1.2.1 of International Atomic Energy Agency document

IAEA-TECDOC403. This IAEA document gives additional guidance and

references IAEA-TECDOC-348, which contains guidance on the seismic design of

structures, systems, and components.

With regard to the allowed dose commitments for seismic events and designs in

Section 3.1 of ANSI/ANS 15.7, the terms "site boundary," "rural zone," and

"urban boundary" are used. For most NRC-licensed non-power reactors, rural

zone" should not be used and "urban boundary" should be assumed to begin at the

"site boundary." However, given applicable site characteristics and emergency

preparedness requirements, the criteria as specified in Section 3.1 of

ANSI/ANS 15.7 could be used.

The above guidance is applicable to research reactors licensed by NRC. For test

reactors the requirements of 10 CFR Part 100 must be applied. The guidance and

criteria of 10 CFR Part 100 are complete and are adequate for assessing test

reactors.

To verify that seismic design functions are met, the applicant should give the bases

for technical specifications necessary to ensure operability, testing, and inspection

of associated systems, including instrumentation and control portions, if applicable.

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3.5 Systems and Components

In this section the applicant should give the design bases for the systems and

components required to function for safe reactor operation and shutdown. For

non-power reactors, this section shouId include, at a minimum, the7 fuel system,

control rod'scram systems, other protective and safety systems; and the

electromechanical systems and components associated with emergency core

cooling systems, reactor room ventilation, confinement or containment systems,

and other systems that may be required to'prevent uncontrolled release of

radioactive material. The design criteria should include the conditions that are

important for reliable operation of the systems and components (e.g., dynamic and

static loads, number of cyclic loads, vibration, wear, friction, strength of materials,

and effects of radiation and tenperature).' The specific application of these design

criteria should generally be given in other chapters of the 'SAR. For example, if

this chapter establishes that a design criterion for the control rods is that it drop by

the force of gravity, Chapter 4, 'Reactor Description," should describe the

electromechanical and reactor dynamic design bases to accomplish this insertion

within a specified time, normally I second.

3.6 References

American National Standards Institute, ANSI N323, 'Radiation Protection

Instrumentation Test and Calibration," ANS, LaGrange Park, Illinois, 1978.

American National Standards Institute/American Nuclear Society,

ANSI/ANS 15.2, 'Quality Control for Plate-Type Uranium-Aluminum Fuel

Elements," ANS, LaGrange Park, Illinois, 1990.

American National Standards Institute/American Nuclear Society,

ANSVANS 15.7, "Research Reactor Site Evaluation," ANS, LaGrange Park,

Illinois, 1977.

American National Standards Institute/American Nuclear Society,

ANSI/ANS 15.8, "Quality Assurance Program Requirements for Research

Reactors," ANS, LaGrange Park, Illinois, 1986.

American National Standards Institute/American Nuclear Society,

ANSI/ANS 15.11, "Radiological Controls at Research Reactors," ANS, LaGrange

Park, Illinois, 1993.

American National Standards Institute/American Nuclear Society,

ANSVIANS 15.15, "Criteria for Reactor Safety Systems for Research Reactors,"

ANS, LaGrange Park, Illinois, 1978.

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. A.-

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American National Standards Institute/American Nuclear Society,

ANSI/ANS 15.17, "Fire Protection Program Criteria for Research Reactors,"

ANS, LaGrange Park, Illinois, 1981.

American National Standards Institute/American Nuclear Society,

ANSI/ANS 15.20, "Criteria for the Reactor Control and Safety Systems of

Research Reactors," (draft), ANS, LaGrange Park, Illinois.

International Atomic Energy Agency, IAEA-TECDOC-348, "Earthquake

Resistant Design of Nuclear Facilities With Limited Radioactive Inventory,"

Vienna, Austria, 1985

International Atomic Energy Agency, IAEA-TECDOC-403, "Siting for Research

Reactors," Vienna, Austria, 1987.

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NUREG- 1 537. PART I 3-6 Rev.0, 2/96
4 REACTOR DESCRIPTION

In this chapter of the SAR, the applicant should discuss and describe the principal

features, operating characteristics, and parameters of the reactor. The analysis in

this chapter should support the conclusion that the reactor is conservatively

designed for safe operation and shutdown under all credible operating conditions.

Information in this chapter of the SAR should provide the design bases for many

systems, subsystems, and functions discussed elsewhere in the SAR and for many

technical specifications.

4.1 Summary Description

In this section the applicant should briefly summarize the design and functional

characteristics of the reactor. The applicant should present the principal safety

considerations in the selection of the reactor type as well as the design principles

for the components and systems that address those considerations. This section

should contain summary tables of important reactor parameters and sufficient

drawings and schematic diagrams to explain and illustrate the main reactor design

features.

The applicant should briefly address the following features of the reactor:

* thermal power level

* fuel type and enrichment

* pool or tank type

* forced aid/or natural-co wection cooling

* type of coolant, moderator, and reflector

* principal features for experimental programs

* pulsing or steady power

* novel concepts requiring substantial new development

4.2 Reactor-Core

In this section the applicant should present all design information and analyses

necessary to demonstrate that the core can be safely operated. The major core

components to be described are fuel, neutron moderator, neutron reflector, control

elements, neutron startup source, incore cooling components,'and any incore

experimental facilities. -The source or basis of the information presented should be

given.

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4.2.1 Reactor Fuel

In this section the applicant should describe the reactor fuel system. Included

should be the design features selected to ensure that the fuel and cladding can

withstand all credible environmental and irradiation conditions during their life

cycle at the reactor site. The discussions should address the incore fuel operating

conditions. Handling, transport, and storage of fuel should be discussed in

Chapter 9, "Auxiliary Systems," of the SAR. Drawings and tables of design

specifications and operating characteristics of the fuel should be presented.

Most non-power reactors contain heterogenous fuel elements consisting of rods,

plates, or pins, which are addressed in the following sections. Homogeneous fuels

should be described and analyzed in a comparable way. Information should be

current; supported by referenced tests, measurements, and operating experience;

and compared with additional applicant experience where applicable. The

information should include the following:

* Chemical composition, enrichment, uranium loading, and important

metallurgical features of the fissile material in the basic fuel unit. The

information should indicate dispersion, alloy, cermet, sintering, and such

special properties as burnable poisons or neutron moderators.

* Description of the basic fuel unit, including plates, rods, pins, or pellets.

This information should include dimensions, fabrication methods, and

cladding or encapsulation methods. Special features, such as moderators

or reflectors, external geometrical designs to enhance cooling capability,

and inherent safety or feedback provisions should be discussed.

* Material and structural information such as dimensions, spacings,

fabrication methods, compatibility of materials, and specifications with

tolerances. All types of fuel elements to be used should be described,

including full elements, partial elements, control rod elements, instrumented

elements, and special elements for experimental facilities. Features that

ensure accurate and secure positioning and adequate cool nt flow should

be described.

* Information on material parameters that could affect fuel integrity, such as

melting, softening, or blistering temperatures; corrosion; erosion; and

mechanical factors, such as swelling, bending, twisting, compression, and

shearing.

* Physical properties with significance in regard to safety and fuel integrity

that are important for the thermal-hydraulic analyses, such as heat capacity,

thennal conductivity, gas evolution or diffusion, occluded or encapsulated

NUREG. 1537. PT I 4-2 REv. 0.2196

NUREG- 1537, PART I 4-2 REV. O. 2196
REACTORDESCRWnON

void volume, fuel burnup limits, capability to retain fission products,

swelling resistance, and buildup of oxides.

* If the reactor is designed for pulsing, any special attributes of the fuel that

contribute to pulsing safety.

* Abriefhistory of the fuel type, with references to the fuel development

program, including summaries of performance tests, qualification, and

operating history. A brief history of the actual fuel elements to be used,

including fabrication, previous irradiation conditions, storage environments,

surveillance procedures, and qualification tests.

* Mechanical forces and stresses, hydraulic forces, thermal changes and

temperature gradients, internal pressures including that from fission

products and gas evolution, and radiation effects including the maximum

fission densities and fission rates that the fuel units and elements are

designed to accommodate.

Limits on operating conditions for the fuel should be supported by information and

analyses. These limits are specified to ensure that the integrity of the fuel elements

and their cladding or fission product barrier will not be impaired. They should

form the design bases for this and other chapters of the SAR, for the reactor safety

limits, and for other fuel-related technical specifications.

4.2.2 Control Rods

In this section the applicant should give information on the control rods, including

all rods or control elements that are designed to change reactivity during reactor

operation. The physical, Idnetic, and electromechanical features demonstrating

-that the rods can fulfill their control and safety functions should be described.

Results of computing control rod reactivity worths may be presented in this

section, but details of the calculation of reactivity effects should appear in

Section 4.5, 'Nuclear Design," of the SAR. The information in this section should

include the following:

* The number and types of rods (e.g., shim, safety, regulating, transient),

their designed locations in the core, and their designed reactivity worths.

The considerations and bases for redundancy and diversity should be

provided. Limits on core configuration should be discussed. :

* The structural and geometric description, including the shape, size,

materials, cladding, fabrication methods, and specifications with tolerances

for the rods. This should include the type and concentration of neutron

absorber, or emitter, if applicable. Also, calculations of changes in

REv. 0,2196 4.3 STANDARD FORMAT AND Co'41tNT

REV.0,2196 - 4-3 STANDARD FORMAT AND CoNTENT
CHAPTER 4

reactivity worth due to burnup and assessment of radiation damage, heating

effects, and chemical compatibility with the coolant and other core

components should be given. If the control rods have followers, the

design, composition, and reactivity effects of the follower should be

discussed.

The design of mechanical supports for the active component, the method of

indicating and ensuring reproducible positioning in the core, and the drive

mechanism of each type of rod. This information should include the source

of motive power, usually electrical, and the systems ensuring scram

capability. For a reactor designed for pulsing, the transient rod should be

described in detail, including its drive mechanisms and the methods for

calibration, pulse reproducibility, and prevention of inadvertent pulsing.

* The kinetic behavior of the rods, showing either the positive or negative

rate of reactivity change, in the normal drive and scram modes of

operation. This information should be supplied for all rods, including

transient rods in a reactor designed for pulsing. The applicant should show

that the control rod design conforms with the shutdown margin

requirements.

* The scram logic and circuitry, interlocks and inhibits on rod withdrawal,

trip release and insertion times, and trip or scram initiation systems should

be summarized here and described in detail in Chapter 7, "Instrumentation

and Control Systems."

* Special features of the control rods, their core locations, power sources,

drive or release mechanisms designed to ensure operability and capability

to provide safe reactor operation and shutdown under all conditions during

which operation is required in the safety analysis if there is a single failure

or malfunction in the control system itself Such features may include

mechanisms to limit the speed of rod movement.

* Technical specification requirements for the control rods and their

justification. These are the limiting conditions for operation, surveillance

requirements, and design features as discussed in Chapter 14, 'Technical

Specifications," of this format and content guide.

4.2.3 Neutron Moderator and Reflector

In this section the applicant should discuss the materials and systems designed to

moderate the neutrons within the fuel region and reflect leakage neutrons back

into the fuel region. The information should include the materials, geometries,

designs for changes or replacement, provisions for cooling, radiation damage

NUREG-1537, PART I 4-4 REV. O. 2/96
REACTORDESCRarON

considerations, and provisions for experimental facilities or special uses. Multipleuse systems and features such as moderator coolant, fuel moderator, and reflector

shield should be described. If moderators or reflectors are encapsulated to prevent

contact with coolant, the effect of faiure of the encapsulation should be analyzed.

It should be possible to operate the reactor safely until failed encapsulations are

repaired or replaced. If reactor operations cannot be safely continued, the reactor

should be placed and maintained in a safe condition until encapsulations are

repaired or replaced. Technical specification requirements should be proposed and

justified for the moderator and reflector in accordance with the guidance in

Chapter 14 of this format and content guide. The nuclear design of the moderator

and reflector should be discussed in Section 4.5 of the SAR.

4.2.4 Neutron Startup Source

In this section the applicant should present design information about the neutron

startup source and its holder. The applicant should show that the source will

produce the necessary neutrons to allow a monitored startup with the reactor

instrumentation. The information should include the neutron strength and

spectrum, source type and materials, its bumup and decay lifetime, and its

regeneration characteristics. Other necessary information includes the material and

geometry of the holder, the method of positioning the source in the core, and the

core locations in which the source is designed to be used. Utilization information

and such limitations as radiation heating or damage and chemical compatibility

with coolant and other core components should be discussed. Any technical

specification limits on the source, such as the maximum power level the reactor

can be run with the source in place (for plutonium-beryllium sources and other

source types that can act as fuel), or surveillance requirements to ensure source

integrity should be proposed and justified in this section of the SAR in accordance

with the guidance in Chapter 14 of this format and content guide.

4.2.5 Core Support Structure

In this section the applicant should present design information about the

mechanical structures that support and position the core and its components. The

information should include the following:,.

The design considerations that ensure that all necessary loads and hydraulic

forces can be conservatively supported with and without the buoyant forces

of the reactor water.

The methods by which core components are accurately and reproducibly

positioned and secured, including specification tolerances, as well as

features of the grid plate such as fuel holddown grids, fuel element spacers,

and control rod guides and supports.

REV. 0,2196 4-5 STANDARD FORMAT AND CONIENT

REV. O. 2/9 4-5 STANDARD FORMAT AND CONTENT
- a

CHAPTER 4

* The materials of construction, including considerations for radiation

damage, corrosion, erosion, chemical compatibility with coolant and core

components, potential effects on reactivity, induced radioactivities, and

maintenance.

* Design features of the core support structure that accommodate other

systems and components such as radiation shields, beam ports or other

experimental facilities, coolant pipes, coolant plenums or deflectors, and

nuclear detectors.

* For a movable core support, design information describing the motive

power system, the system for ensuring position, and interlocks that prevent

or control motion while the reactor is critical, while forced cooling is

required, or while other activities that prohibit core support movement are

to be conducted, if such a system is required (e.g., experimental facility

operations).

* Technical specifications that control important design features, limiting

conditions for operation, and surveillances as discussed in Chapter 14 of

this format and content guide. The applicant should justify these technical

specifications in this section of the SARI

4.3 Reactor Tank or Pool

The cores of most non-power reactors are immersed in water within a tank or'

pool. In this section the applicant should present all information about the tank or

pool necessary to ensure its integrity. The information should include the

following:

* Design and considerations to ensure that no hydrodynamic, hydrostatic,

mechanical chemical, and radiation forces or stresses could cause failure or

loss of integrity of the tank during its projected lifetime over the range of

design characteristics.

* Design and dimensions to ensure sufficient shielding water to protect

personnel and components, as well as sufficient depth to ensure necessary

coolant flow and pressures. (Also see Sections 4.4 and 4.6 and

Chapter 11, 'Radiation Protection Program and Waste Management," of

this format and content guide.)

Designs and description of materials, including dimensions, supporting

structures, chemical compatibility with the coolant and other reactor

system components, radiation fields and any consequences of radiation

NUREG-1537,PART I 4-6 REV 0, 2/96
REACroR DESOqmoN

damage, protection from corrosion in inaccessible regions, and capability to

replace components, if necessary.

Locations of penetrations and attachment methods for other components

and pipes. The relationships of these penetrations to core and water

surface elevations should be discuised. Safety-related features that prevent

loss of coolant should be discussed and related to Sections 4.4 and 4.6 and

to the loss-of-coolant-accident scenarios analyzed in Chapter 13, "Accident

Analyses," as applicable.

Planned methods for assessing radiation damage, cheniiWaI damage, or

deterioration during the projected lifetime. In this section the applicant

should assess the possibility of uncontrolled leakage of contaminated

primary coolant and should discuss preventive and protective features.

* Technical specifications that control important design features, limiting

conditions for operation, and surveillance requirements as discussed in

Chapter 14 of this format and content guide. The applicant should justify

these technical specifications in this section of the SAR

4.4 Biological Shield

In this section the applicant should present information about the principal

biological shielding designed for the reactor. The information should include the

following:

* The design bases for the radiation shields (e.g., water, concrete, or lead),

including the projected reactor power levels and related source terms and

the criteria for determining the required protection factors 'for all applicable

nuclear radiation activity. Information about'conformiance with the

regulations for radiation exposure and the facility ALARA (as low as is

reasonably achievable) program should appear in Chapter 11. The design

basis should include the designed reactor power levels, the associated

radiation source terms, and other radiation sources within the pool or tank

that require shielding.

* The design details and the methods used to achieve the design bases.' The

applicant should discuss the protection-of personnel and equipment'

finictions. The information should specify the general size and shape of the

shields and the methods used to ensure'structural strength, rigidity, and

functional integrity. The applicant should discuss' the distribution of

shielding factors between liquid (water) and solid (concrete, lead, etc.)

materials. If loss of shield integrity could cause a loss-of-coolant accident,

the features to prevent the loss of integrity should be described.

REv. 0.2196 4-7 STANDARD Fo1wT AND CoIENT

REV. O. 2/96 -4-7 STANDARD FORMAT AND CoNTer
CHAPT4

The materials used and their shielding coefficients and factors, including a

detailed list of constituents and their nuclear and shielding properties. The

applicant should discuss radiation damage and heating or material

dissociation during the projected lifetime of the reactor, induced

radioactivity in structural components; potential radiation leakage or

streaming at penetrations, interfaces, and other voids; shielding at

experimental facilities; and shielding for facilities that store fuel and other

radioactive materials within the reactor pool or tank.

* The assumptions and methods used to calculate the shielding factors,

including references to and justification of the methods. Detailed results of

the shielding calculations should give both neutron and gamma-ray dose

rates at all locations that could be occupied. The applicant should calculate

shield penetrations and voids, such as beam ports, thermal columns, and

irradiation rooms or vaults, as well as the shielding of piping and other

components that could contain radioactive materials or allow radiation

streaming.

* Methods used to prevent neutron irradiation and activation of ground

water or soils surrounding the reactor shield that could enter the

unrestricted environment. The applicant should estimate the maximum

activity should such activation occur and describe remedial actions.

* Technical specifications that control important design features, limiting

conditions for operation, and surveillance requirements as discussed in

Chapter 14 of this format and content guide. The applicant should justify'

these technical specifications in this section of the SAR.

Regulatory Guide 2.1, "Shield Test Program for Evaluation of Installed Biological

Shielding in Research and Training Reactors" is given as Appendix 4. 1.

4.5 Nuclear Design

In this section the applicant should give information on the nuclear parameters and

characteristics of the reactor core and should analyze the kinetic behavior of the

reactor for steady-state and transient operation throughout its life cycle of allowed

cores and burnup as discussed in the safety analysis. The descriptions, analyses,

and results should address all safety issues in the design and operation of the

reactor and should support the conclusion that the reactor can be built and

operated without unacceptable risk to the health and safety of the public. A

detailed description of the analytical methods used in the nuclear design should be

given. Computer codes that are used should be described in detail as to the name

and type of code, the way it is used, and its validity on the basis of experiments or

confirmed predictions of operating non-power reactors. Code descriptions should

NUREG-1537,FART I 4-8 REv.O.2196

NUREG-1537,PARY 1 4-8 RE-V. O. 2196
REAcroR D~sc~noN

include methods of obtaining parameters such as cross sections. Estimates of the

accuracy of the analytical methods should be included. Tables and figures should

be used as necessary to present information clearly.

4.5.1 Normal Operating Conditions

In this section the applicant should present information on the core geometry and

configurations. Operating core configurations should be compact, with no

vacancy in which fuel could be inserted within the core periphery. The limiting

core configuration for a reactor is the core that would yield the highest power

density using the fuel specified for the reactor. All other core configurations

should be demonstrated to be encompassed by the safety analysis of the limitiixg

core configuration. Further information on power density limitations should be

given in Sections 4.5.3 and 4.6. The information in the SAR should include the

following:

* The number, types, and locations of all core components on the grid plate,

including fuel, control rods, neutron reflectors, moderators, incore

experimental components, and core-associated cooling components. If this

information appears elsewhere in the SAR, the section where it is located

should be referenced.

* Descriptions of planned core configurations during the life of the reactor,

showing how a compact core is ensured.

* Discussions and analyses of the reactor operating characteristics. The

applicant should give in detail the effects of changes in configuration and

fuel burnup. If applicable, the applicant should analyze safety-related

considerations for all requested operating modes (e.g., steady power and

pulsing).

* Changes in core reactivity with fuel burnup, plutonium buildup, and

poisons, both fission products and those added by design, if applicable.

* Analyses of the reactor Idnetic behavior and the design requirements and

dynamic features of the control rods that allow controlled operation for all

possible reactor conditions.

* Analyses of the basic reactor criticality physics, including the interacting

effects of fuel, neutron moderators and reflectors, control rods, and incore

or in-reflector components such as experimental facilities.

REV. 0, 2196 4-9 REV.,219 FORMAT 4.9STANDARD AM CONTrENT
CHAPTER 4

* Discussion of the safety considerations for different core configurations,

including a limiting core configuration that would yield the highest power

densities and fuel temperatures achievable with the planned fuel.

* The individual reactivity worths of fuel elements, reflector components,

incore and in-reflector components, experimental components, and control

rods in allowed positions. If experimental facilities or components could be

voided or flooded, the reactivity effects and safety considerations should be

included.

* The calculated core reactivities for all core configurations. including the

limiting configuration that would yield the highest possible power density.

* Discussion of the administrative and physical constraints to prevent

inadvertent addition of positive reactivity.

* Technical specifications that control important design features, limiting

conditions for operation, and surveillance requirements as discussed in

Chapter 14 of this format and content guide. The applicant should justify

these technical specifications in this section of the SAP

4.5.2 Reactor Core Physics Parameters

In this section the applicant should discuss the core physics parameters and show

the methods and analyses used to determine them. The information should include

the following:

* Analysis methods and values for neutron lifetime and effective delayed

neutron fraction. The applicant should describe the effects of reactor

operating characteristics and fuel burnup.

* Analysis methods, values, and signs for coefficients of reactivity (e.g., fuel

and moderator temperature, void, and power). The applicant should

describe the effects of reactor operating characteristics and fuel burnup.

This analysis, along with the analysis in Chapter 13, should show that

reactivity coefficients are sufficiently negative to prevent or mitigate

damaging reactor transients.

* The axial and radial distributions of neutron flux densities, justifications for

the methods used, and comparisons with applicable measurements. The

applicant should describe changes in flux densities with power level, fuel

burnup, core configurations, and control rod positions. The information on

neutron flux density should include peak-to-average values for thermalhydraulic analyses. The applicant should validate these calculations by

NUREG-I537,PJT I 4-10 REV. 0, 2./96

NUREG-1537, PART I 4-10 REV. O. 2/96
REACrOR DESnON

comparing them with experimental measurements and other validated

calculations.

Technical specifications that control important design features, limiting

conditions for operation, and surveillance requirements as discussed in

Chapter 14 of this format and content guide. The applicant should justify

these technical specifications in this section of the SAL

4.5.3 Operating Limits

The applicant should present the following information on reactor operating limits:

* Reactivity conditions, excess reactivity, and negative reactivity for

combinations of control rods inserted that are analyzed for the limiting core

and operating cores during the life of the reactor. -.The applicant should

discuss operational and safety considerations for excess reactivity.

* The excess reactivity based on reactor temperature coefficients, poisons,

and experiment worths. The applicant should justify the upper limit on

excess reactivity to ensure safe reactor operation and shutdown.

* The amount of negative reactivity that must be available by control rod*

action to ensure that the reactor can be shut down safely from any

operating condition and maintained in a safe shutdown state. The analyses

should assume that the most reactive control rod is fully withdrawn (one

stuck rod), non-scrammable control rods are at their most reactive

position, and normal electrical power is unavailable to the reactor. The

applicant should discuss how shutdown margin will be verified. The

iialyses should include all relevant uncertainties and error limits.

* The limiting core configuration that is possible with the planned fuel in this

reactor. The limit should be imposed by the maximum neutron flux density

and thermal power density compatible with coolant availability. The safety

limits and limiting safety system settings for the reactor should be derived

from this core configuration. The detailed analyses should be included in

Section 4.6. Normal operating conditions and credible events, such as a

stuck control rod, should be considered.

* A transient analysis assuming that an instrumentation malfunction drives

the most reactive control rod out in a continuous ramp mode in its most

reactive region. This analysis can also be based on a credible failure of a

movable experiment. It should show that the reactor is not damaged and

fuel integrity is not lost.

REV. 0, 2196 4-li STANDARD FORMAT 4 u, CONIENT

REV. 0.2/96 4-11 STANDARD FORMAT ASD Comerr
.

CHAPTER 4

* The redundancy and diversity of control rods necessary to ensure reactor

control for the considerations noted above.

* Technical specifications for safety limits, limiting safety system settings,

limiting conditions for operation, and surveillance requirements as

discussed in Chapter 14 of this format and content guide. The applicant

should justify these technical specifications in this section of the SAR

4.6 Thermal-Hydraulic Design

In this section the applicant should present the information and analyses necessary

to show that sufficient cooling capacity exists to prevent fuel overheating and loss

of integrity for all anticipated reactor operating conditions, including pulsing, if

applicable. The applicant should address the coolant flow conditions for which the

reactor is designed and licensed, forced or natural-convection flow, or both. A

detailed description of the analytical methods used in the thermal-hydraulic design

should be provided. Computer codes that are used should be described in detail as

to the name and type of code, the way it is used, and its validity on the basis of

experiments or confirmed predictions of operating non-power reactors. Estimates

of the accuracy of the analytical methods should be included. The information

should include the following:

* The coolant hydraulic characteristics of the core, including flow rates,

pressures, pressure changes at channel exits and entrances, and frictional

and buoyant forces. The applicant should address individual heated

channels as well as the core as a whole for all flow conditions in the

primary coolant system. The transition from forced to natural-convection

flow for all forced-flow reactors should be calculated, and the applicant

should prepare calculations for an event during which normal electrical

power is lost.

* The thermal power density distribution in the basic fuel units and heat

fluxes into the coolant of each channel and along the channel, derived from

the fuel loading and neutron flux characteristics discussed above.

* Calculations and the thermal-hydraulic methodology for the transfer of heat

to the coolant. The applicant should take into account uncertainties in

thermal-hydraulic and nuclear parameters and such engineering factors as

plate thickness, gap width, and the buildup of cladding oxides. The

calculations should be based on fuel measurements and procurement

specifications, as well as operating history and conditions. The

calculational methodology should be applicable to the thermal-hydraulic

operating conditions, and the applicant should justify its use.

NLJREG-1537, PART I 4-12 REv. 0, 2/96

NUREG I1537, PART I 4-12 REV. O. 2/96
REAcTDEsauPfnoN

The calculations and experimental measurements to determine the coolant

conditions ensuring that fuel and cladding integrity are not lost. The

applicant should calculate at least the limiting core configuration.

Operating conditions should include steady fission power, shutdown decay

heat, planned pulses, and transients analyzed in Chapter 13. The applicant

should take into account operational and fuel characteristics from the

beginning to the end of fuel life.

* For the core geometry and the coolant thermal-hydraulic characteristics, a

discussion to establish the fuel heat removal conditions that ensure fuel

integrity such as fuel surface saturation temperature, onset of nucleate

boiling, departure from nucleate boiling and/or flow instability. The

discussion should show correlations among these factors and justify their

use in deriving safety limits and limiting safety system settings for the

technical specifications.

* The design bases for the primary coolant system, emergency core cooling

system, and other systems designed to maintain fuel integrity, which should

also be discussed in Chapter 5, 'Reactor Coolant Systems." The analyses

here and in Chapter 13 should describe loss-of-coolant scenarios for

forced-flow reactors. Natural-convection cooling that removes decay heat

to ensure thermal stability should also be discussed. Flow blockages

should be analyzed in Chapter 13.

* Detailed analyses for a pulsing reactor containing descriptions of the core

configurations; the bases of the feedback coefficients; the calculational

model and assumptions; the thermal-hydraulic evolution during a pulse;

core, transient rod, and fuel characteristics that determine the shape and

magnitude of a pulse; and the safety considerations that establish limits to

pulse sizes. Any changes in fuel parameters resulting from steady-power

operation that could affect pulse characteristics should be analyzed. These

changes could include burnup, hydrogen migration, cladding oxidation, and

decrease in burnable poison, as applicable. The analyses should form the

bases for technical specifications that limit reactor operating conditions,

process variables, and pulse rod reactivity worths.

REV. 0, 2196 4-13 STANDARD FORMAT AND CONTE

S
Appendix 4.1

Regulatory Guide 2.1

Shield Test Program for

Evaluation of Installed Biological Shielding

in Research and Training Reactors

STA}ARD FORMAT Am CoNYEN
MaY 1973

U.S. ATOMIC ENERGY COMMISSION

REGULATORY GUIDE

DIRECTORATE OF REGUWATORY STANDARDS

REGULATORY GUIDE 2.1

SHIELD TEST PROGRAM FOR

EVALUATION OF INSTALLED BIOLOGICAL SHIELDING

IN RESEARCH AND TRAINING REACTORS

A. INTRODUCTION

Subdivision (bX6) (NO) of section 5034. "Content%

of applications technical mnformation." or IOCFR Part

50. 'Licensing of Production and Utlitation Factlities,"

Ecquires an applicant for a license to include an his final

safety analysis report plans for preoperational testing

and initial operation. This regulatory guide describes a

shield test program that b generally acccptable for

evaluation of installed biological shielding in researmh

and traming reactors

B. DISCUSSION

Subcomnmittec ANS 6, Shielding, of the AmerLan

Nuclear Society Standards Committee hi developed j

standard that describes an operational shield test

program which may be used in evaluating the ii~stallcd

biological shielding in reseirch and training reaLtors.

This standard was approved by the Amenran National

Standards Committee N18. Nudlear Design Criteria uand

its Secretariat. It was subsequently approved and

designated ANSI N18.9-197' by the American National

Standards Institute (ANSI) on September 15. 197'

C. REGULATORY POSITION

The requirements and guidelines contained in ANSI

Nl8 .91972. 'Program for Testing Biologial Shielding

in Nuclear Reactor Plant."' Approved September 15.

1972. are generally .cceptAhle And. with due

consideration for the unique ciutrateristici of eJII

research and titning rcji.t. provide An Adequate b iti

for conducting a shicId lest propram during

preoperational and 1tarlup testing for evaluation of

'Copies may be obtiasId lroii Amkian Nustckr SoLiety. 244 East Orden AvWeue. Iin'datc. 11inoi%6lt5Zl

instaUed biological shielding in

reactors subject to the following.

research and training

1. Section 3214 of ANSI 18 9.1972 defines accessible

areas, controllcd areas, and unlimited access areat

Section 3.2 5 defines Maximum Permissible Dose rate.

Nothing in these paragraphs should imply that exposures

need not be controlled to the requirements of 10 CFR

Part 20, 'Standards for Protection Aainst Radiation."

2. Section S 2 of ANSI 18 9-1972 states that

pro edures for implementing the minimum shield test

program shall be prepared. These procedures should be

designed so that exposures to personnel performing the

test program are as low as practicable. These procedures

should also be designed so that safety hazards to

personnel performing the shield test program are

properly identified For example, gas iotonzgshould

be requured where gases or vapors could affect the

aceessibility of an area.

3 Se tion 6 of ANSI N18.9-1972 speafies tests that

should be conducted for evaluation of installed

biological shielding. This section further specifies use of

survey meters when conducting the required tests. The

shield test program should also include provstons for

gamma and neutron film mapping of critical areas whew

personnel expouure may occur due to streaming, cracks.

or gaps in the shielding too small to detect by survey

meters. e g areas in the vicinity of beam holes.

irradiation ports, or shielding areas directly aligned with

the Lore

4 Section 9 of ANSI N18 9-1972 states that

instiument, used In carrymng out the minmum shield

test program shall have been calibrated prior to use in

the test progrnm and immediately after each survey The

shield test program should also include provisions for

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&librating all radiatit surnvy monitors (both portable

and Instalcd) aplat a wurm emritfig radiation of

appeozimately the same type and intensity as that

expected to be measured durinng h survey.

S. Sections 92 and 9.3 of ANSI 18.9-1972 provide

requirements for survey irntruments. In addition to

thse requiremcnts. a survey instrument's range should

be consistent with the actual dose rump expected. For

messurements conducted while a reactor is operating in

the puAbd mode. appropriate instrumentation such as

ram packets, which wNl properly respond to and

measure radition during the pulsed mode of operation

should be provided.

2.1-2
5 REACTOR COOLANT SYSTEMS

In this chapter of the SAR, the applicant should give the design bases, descriptions,

and functional analyses of the reactor coolant systems. The principal purpose of

the coolant system is to safely remove the fission and decay heat from the fuel and

dissipate it to the environment. The discussions should include all significant heat

sources in the reactor and should show how the heat is safely removed and

transferred to the environment.

The coolant in the primary systems of most non-power reactors serves more

functions than just efficient removal of heat. It can act as a radiation shield for the

reactor, fuel storage facilities, and, in some designs, experimental facilities and

experiments. In open-pool reactors, the coolant is the only vertical shielding. In

many designs the reactor coolant also'acts as a core moderator and reflector.

Because of these many functions of the reactor coolant, the design of the reactor

coolant systems is based on selecting among interdependent parameters, including

thermal power level, research capability, available fuel type, reactor core physics

requirements, and radiation shielding.

Some non-power reactors are licensed to operate at such low power levels that no

significant temperature increases will occur during normal operation. Such

reactors may not require an engineered coolant system'for heat removal.' For those

reactors, the applicant should, in Chapter 4, 'Reactor Description," of the SAR,

discuss the disposition of the heat produced, estimate potential temperature

increases during operation, and justify why an engineered coolant system for heat

removal is not required. In this chapter the applicant'should summize those

considerations and conclusions.

For all other non-power reactors, the applicant should descnbe and discuss in this

chapter systems to remove and dispose of the waste heat. The design bases of the'

reactor coolant systems for the full range of normal operation should be based on

ensuring acceptable reactor conditions established in Chapter 4 of the SAR. The

design bases of any features of the core cooling system designed to respond to

potential accidents or to mitigate the consequences of potential accidents should be

derived from the analyses in Chapter 13, "Accident Analyses." These features'

should be summarized in this chapter and discussed in detail in Chapter 6,-'

"Engineered Safety Features." In this chapter the applicant should discuss and

reference the technical specifications where analyses are used as the basis for a

requirement.

In this chapter the applicant should describe all auxiliary and subsystems that use

and contribute to the heat load of either the primary or secondary coolant system.

Any auxiliary systems using coolant from other sources should be discussed in

Chapter 9, 'Auxiliary Systems."

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The primary loops of the coolant systems of most licensed non-power reactors are

of two basic types, forced-convection and natural thermal-convection. Facilities

using forced-convection cooling also may be licensed to operate in the naturalconvection mode. All non-power reactors with engineered coolant systems that do

not have active decay heat removal systems should be capable of dissipating decay

heat in the natural-convection mode. In this chapter the applicant should describe

the complete coolant systems for the allowed modes of operation, as discussed

below.

5.1 Summary Description

In this section the applicant should give a brief description of the reactor coolant

systems, summarizing the principal features. Information should include the

following:

* type of primary coolant: liquid, gas, or solid (conduction to surrounding

structures)

* type of primary coolant system: open or closed to the atmosphere

e type of coolant flow in the primary coolant system: forced-convection,

natural-convection, or both

* type of secondary coolant system, if one is present, and the method of heat

disposal to the environment

* capability to provide sufficient beat removal to support continuous

operation at full licensed power

* special or facility-unique features

5.2 Primary Coolant System

The basic requirements and design bases of the primary coolant system are to

maintain reactor facility conditions within the range of design conditions and

accident analyses assumptions derived from other chapters of the SAR, especially

Chapters 4 and 13. The applicant should show the interrelationships among all

SAR chapters and the way the designed primary coolant system provides all

necessary functions. The following information should be included:

* Design bases and functional requirements of the primary coolant system.

* Schematic and flow diagrams of the system, showing such essential

components as the heat source (reactor core), heat sink (heat exchanger),

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pumps, piping, valves, control and safety instrumentation, interlocks, and

other related subsystems.

Tables of allowable ranges of important design and operating parameters

and specifications for the primary coolant system and its components, including

- coolant material

- coolant flow rates

-- inlet and outlet tempeitures and pressu throughout the system

- - elevation of components and water levels relative to the reactor

core

- construction materials of components

fabrication specifications of saety-related components

- coolant quality requirements for operation and shutdown

conditions, including pH and conductivity at a minimum

- minimum coolant level

Discussions and analyses keyed to drawings showing how the system

provides the necessary cooling for all heat loads and all potential reactor

conditions analyzed in the thermal-hydraulics'section of Chapters 4 and 13,

including the following:

- Removal of heat from the fuel by forced-convection or naturalconvection cooling, or both for those reactors licensed to operate in

both modes. Discussion and analyses of the effect of the size,

shape, and structural features of the primary vesse or pool on

cooling characteristics; the function of the pool as a heat reservoir,

and the effect of water depth on nabural thermal convection cooling.

Transfer of heat from the primary coolant to 'a secondary coolant

' system for all reactor conditions. This discussion should include

-any heat exchanger design and operating conditio'ns. 'Some nonpower reactors may have only a primary'coolant 'system'that

functions as a heat reservoir. For such systems, the analyses should

include any fiacors that limit continuous operation, such as pool

water temperature, and the proposed technical specifications that

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ensure operation within the analyzed limits. Some high-power nonpower reactors may require shutdown pumps to circulate coolant

through the core to remove decay heat.

- Safe reactor shutdown, including passive or fail-safe transition from

forced- to natural-convection cooling and removal of decay heat

from the fuel. This discussion should include the loss of offsite

electrical power.

- Locations, designs, and functions of such essential components as

drains, syphon-breaks, pumps, isolation valves, and check valves.

These components ensure that the primary coolant system is

operable and that uncontrolled loss or discharge of contaminated

coolant from the primary system does not occur. Radiological

effects of potential coolant releases should primarily be analyzed in

Chapter 11, 'Radiological Protection Program and Waste

Management."

* Discussion of the control and safety instrumentation, including location and

functions of sensors and readout devices. The scram or interlock functions

that prevent safety limits from being exceeded should be shown and

discussed, including the related technical specifications.

* Description and function of any special features of the primary coolant

system, such as removal of the neutron moderator for backup reactor

shutdown.

* Brief description and functions of special features or components of the

primary coolant system that affect or limit personnel radiation exposures

from such radionuclides as nitrogen-16 and argon-41 and from radioactive

contaminants and fission products.

* Description of radiation monitors or detectors incorporated into the

primary coolant system and discussion of their functions.

* Brief discussion and references to detailed discussions in later sections of

auxiliary systems using primary coolant, such as coolant cleanup, makeup

water, nitrogen-16 control, emergency core cooling, experiment cooling,

experimental facility cooling, and biological or thermal shield cooling. The

direct effect of these auxiliary systems on the design and functioning of the

primary coolant system should be discussed.

* Brief discussion of radiation shielding provided by the primary coolant.

Most non-power reactors are submerged in a pool or tank so that the

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primary coolant shields personnel above the pool and at the reactor room

floor. The design bases for these shielding functions should be analyzed for

anticipated reactor conditions in Chapters 4 and 11 of the SAR and for

postulated reactor accidents in Chapter 13. The effect of any special

shielding features, such as fuel storage facility shielding and experimental

facility shielding (e.g., beam tubes), on the functioning of the primary

coolant system should be discussed.

* Discussion of leak detection and allowable leakage limis, if any. Prompt

detection of leakage is very important in reactors with heavy water.

* Discussion of normal primary coolant radiation concentration limits,

including sampling frequency, isotopes of interest, and actions to be taken

if limits are exceeded.

* For reactors that have closed systems, a discussion of allowable hydrogen

limits in air spaces that are in contact with the primary coolant.

* Discussion of technical specification requirements for parameters of the

primary coolant system, including the bases and surveillance requirements.

5.3 Secondary Coolant System

Many licensed non-power reactors include fuel cooling systems composed of both

primary and secondary coolant systems. Some very low powered reactors contain

no engineered coolant system, and some may contain a single component (primary)

coolant system. Still other low-powered reactors may include a coolant system,

with both primary and secondary subsystems, that would not support continuous

reactor operation at full licensed power.

In this section the applicant should give information about those non-power

reactors that include a secondary coolant system. For the others, the applicant

should state that a secondary coolant system is not needed and should justify that

conclusion. The following information should be provided:

* The design bases and functional requirements of the secondary coolant

system, including whether the system is designed for continuous full-power

reactor operation and whether it is shared with other facilities.

* Schematic and flow diagrams of the secondary coolant system, showing

such essentials as how the heat exchanger connects the primary coolant

system (the heat source) to the secondary coolant system, pumps, piping,

valves, control and safety instrumentation, interlocks, and interface with

the environment for ultimate release of the heat.

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Tables of the range of important design and operating parameters and

specifications of the secondary coolant system, including the following:

- Coolant material and its source.

- Coolant flow rates.

- Type of heat dissipation system, such as cooling tower, refrigerator,

radiator, or body of water.

- Location of heat dissipation system in relation to the reactor and

the heat exchanger.

- Construction materials and fabrication specifications of

components. (For older facilities for which complete information

may not be available, the applicant should make a best effort to

provide this information and should discuss the operating history of

components.)

- Heat dissipation specifications related to environmental fiactors

(e.g., temperature and humidity).

- Specifications and limitations on coolant quality and corrosion

of the secondary coolant system components including the

environmental effects of the use of secondary coolant chemicals.

* Discussion and functional analyses keyed to the drawings showing how the

system provides the necessary cooling for all potential reactor conditions.

These discussions should address the following:

- Inlet and outlet temperatures and pressures throughout the system,

including the pressure differential between the primary and

secondary coolant systems in the heat exchanger. (The applicant

should discuss how the pressure in the secondary coolant system is

maintained above that in the primary coolant system for all

operating conditions, or analyze the radiological effect of leakage of

contaminated primary coolant into the secondary coolant system.

Isolation of the heat exchanger during shutdown periods is an

acceptable method to control potential primary-to-secondarysystem leakage if secondary coolant system pressure is lower than

primary coolant system pressure only during periods of system

shutdown. The applicant does not need to perform an analysis of

primary-to-secondary-system leakage if secondary coolant system

pressure is lower than primary coolant system pressure for only

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short periods for system testing or repair. If the transfer of primary

coolant into the secondary coolant system is caused by an abrupt

event, such as atube rupture in the heat exchanger, the analysis

should be given in Chapter 13 and summarized here.)

Control of heat removal from the secondary coolant system

necessary to maintain fuel temperatures in the primary coolant

system within the limits derived in the thermal-hydraulics analyses

inChapters 4 and 13 of the SARI

Removal of heat from the heat exchaiger. and release to the

environment when the primary coolant system operates in all

anticipated and licensed modes, including forced-convection flow

and natural-convection flow, as'applicable.

Safe reactor shutdown and removal and dissipation of decay heat,

including evaluation of the primary coolant system change from

forced-convection flow to natral-convection flow if forcedconvection flow is an allowed mode of operation.

Response of the secondary coolant system to the loss of primary

coolant with or without an emergency core cooling system.

- Locations, designs, and functions of such essential components as

drains, sumps, pumps, makeup water, and check valves that ensure

contaminated primary coolant is not inadvertently transferred to the

secondary coolant system and released to the environment.

Discussion of control and safety instrumentation, including locations and

functions of sensors and readout devices and interlocks or safety

capabilities.

* Descriptions of functions of any radiation" monitors or 'detectors

incorporated into the secondary coolant system Discussion of surveillance

to measure secondary coolant activity including frequency, action levels,

and action to be taken.

* Brief comments and reference to detailed discussion in other sections of

auxiliary cooling systems that transfer heat to the 'secondary coolant

system, such as emergency core cooling system, experiment cooling, or

biological or thermal shield cooling.

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Discussion of technical specification requirements, as appropriate, for the

secondary coolant system, including the bases and surveillance

requirements.

5.4 Primary Coolant Cleanup System

Most licensed non-power reactors contain solid fuel elements immersed in the

primary coolant water. Experience has shown that the metal cladding is

susceptible to corrosion if the chemical purity of the water is not high. The water

purity must be above the usual purity of the potable water supply. Experience has

shown that oxide buildup on aluminum-clad fuel operating at high power densities

can reduce heat transfer. The rate of buildup depends, among other factors, on the

water quality (Griess et al., 1964). This process should be evaluated in Chapter 4

of the SAR and summarized in this chapter if it contributes to determining the

requirements for primary coolant purity. To delay or prevent component failure by

corrosion, non-power reactors should have a primary coolant cleanup system. The

purity of the primary coolant should be maintained as high as reasonably possible

for the following reasons:

* to limit the chemical corrosion of fuel cladding, control and safety rod

cladding, reactor vessel or pool, and other essential components in the

primary coolant system

* to limit the concentrations of particulate and dissolved contaminants that

could be made radioactive by neutron irradiation

d to maintain high transparency of the water for observation of submerged

operational and utilization components

The applicant should give the following information:

* The design bases and functional requirements of the primary coolant

cleanup system. Experience at non-power reactors has shown that with a

well-planned water cleanup system and good housekeeping practices,

primary coolant quality can be maintained within the following ranges:

- electrical conductivity sS1 jxmho/cm

- pH between 5.5 and 7.5

The design bases should be derived from discussions in Chapter 4 of the

SAR, and any recommendations of the fuel vendor also should be

addressed.

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* Schematic drawings and flow diagramsof the primary coolant cleanup

loop.

* Table of specifications for the cleanup system demonstrating that it is

* designed for the volume and throughput of the primary coolant system

(For older facilities for which complete information may not be available,

the applicant should make a best effort to provide this information and

should discuss the operating history of components.)

* Locations and functions of control and monitoring instrumentation,

including sensors, recorders, and meters. The discussion of monitors

should include methods 'for continuously assessing coolant quality and

effectiveness of the cleanup system.

* Locations and functional designs of cleanup system components such as

branch points, pumps, valves, filters, and demineralizers.

* Discussion of schedules and 'methods for replacing or regenerating resins

and filters and disposing of resultant radioactivity to ensure that radiation

exposures do not exceed the liriits discussed in Chapter 11 of the SAR

* Summary of methods for predicting, monitoring, and shielding radioactivity

deposited in filters and demineralizers from routine operations. The

detailed discussion should be-in Chapter 11.

* Summary of methods for predicting and limiting exposures of personnel in

the event of inadvertent release of excess radioactivity in the primary

coolant system and deposition in'filters and dernineralizers. The detailed

discussion should be in Chapter 13.

* Provisions in the design and operation of the cleanup system to avoid

malfunctions that could lead to significant loss of primary coolant or

release of cont i coolant, which could cause radiological exposure

of personnel or release to the unrestricted environment to exceed the

requirements in 10 CFR Part 20 and the facility ALARA (as low as is

reasonably achievable) program guidelines.

* Discussion of technical specification requirements for the primary coolant

cleanup system, including the bases and surveillance requirements.

5.5 Primary Coolant Makeup Water System'

During operations at non-power reactors with a water-based primary coolant

system, primary coolant must be replaced or replenished. Coolant may be lost as a

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result of evaporation from open-pool systems, radiolysis, designed leakage as from

pump seals, and other operational activities. The non-power reactor design should

include a system or a procedure that meets the projected needs for coolant. The

makeup water system need not be designed to provide a rapid, total replacement of

the primary coolant inventory, but should be able to maintain the minimum

acceptable water quantity and quality for reactor operation.

The applicant should provide the following information:

* The design bases for the primary coolant makeup water system that

account for all activities that could cause a decrease in the primary coolant.

A large loss-of-coolant event should be analyzed in Chapter 13 of the SAR.

Although a required emergency core cooling system need not be a part of

the makeup water system, it should be discussed in Chapter 6 if it exists.

* Schematic diagrams and functional discussions that show the source of

water, the methods of addition to the primary coolant system, and the

requirements for pretreatment before addition. Not all non-power reactors

need a makeup water system attached to the reactor primary coolant

system.

* Locations and functions of control instrumentation, including sensors,

readout displays, and interlocks. Methods should be discussed for tracking

additions of makeup water to detect significant changes that might indicate

leaks or other malfunction of the primary coolant system.

Discussion of safety systems and administrative controls to ensure that the

system or procedures for adding makeup water will not lead to significant

loss of primary coolant and will prevent leakage of contaminated coolant

into the potable water supply.

* Discussion of technical specification requirements for the primary coolant

makeup water system, including the bases and surveillance requirements.

5.6 Nitrogen-16 Control System

When ordinary oxygen is irradiated with neutrons of sufficient energy, nitrogen-16,

a high-energy beta and gamma emitter with a 7-second half-life, is formed. In

water-cooled reactors operated above a few hundred kilowatts, the radioactivity of

this nuclide may require specific systems or procedures for limiting personnel

exposure.

In reactors with natural-convection cooling, transport of nitrogen-16 to the pool

surface may be delayed by a coolant circulator or diffuser. In reactors with forcedNUREG-1537. PA.RT I 5-10 REv. 0, 2/96

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convection cooling, the coolant carrying the nitrogen-16 out of the core may be

passed through a system such as a large shielded and baffled tank. This delay

allows the radioactivity to decay significantly before the coolant emerges from the

shieldin&. Another method of radiation control is to shield the entire primary

coolant system:

The applicant should analyze the potential for personnel exposure to nitrogen-16

and propose control systems or procedures that include the following:

* Design bases and functional design of the nitrogen-16 control system or

procedures. The design bases should be derived from analyses in

Chapter 11 ofthe SAR.

* Schematic drawings and system and component specifications for the

nitrogen-16 control system.

* The method used by the nitrogen-I 6 control system to reduce exposure

rates and potential doses in occupied areas. Potential doses with and

without the nitrogen-16 controls should be analyzed in Chapter 11 and

summarized in this section of the SAR. These potential doses include dose

from direct radiation and dose from airborne nitrogen-16.

* The effect of the nitrogen-16 control system on overall reactor safety and

operation. For example, diffuser systems in natural convection reactors

may affect coolant flow and coolant transparency.

* Other reactor design features affected by the nitrogen-16 control system

For example, a large shielded decay tank may affect coolant flow

parameters, pump sizes, access for surveillance or inservice testing, or

other factors for the primary coolant system.

* An assessment that the nitrogen-16 control system would not lead to an

uncontrolled loss of primary coolant or the release of contaminated primary

coolant that exceeds the requirements in 10 CFR Part 20 and the facility

ALARA program guidelines. Methods for analyzing radiation exposures as

a result of coolant release should be consistent with the analyses in

Chapter 1 1.

* Discussion of any technical specification requirements for the nitrogen-16

control system, including the bases and surveillance requirements.

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5.7 Auxiliary Systems Using Primary Coolant

In addition to the systems discussed above that are associated with the primary

coolant system, other auxiliary cooling systems or shields in some non-power

reactors may require the use of primary coolant and may affect the operation or

safety of the reactor. Any function of the primary coolant that is principally

shielding should be described in Chapter 4 and summarized in this section and

Chapter 11 of the SAR. If the reactor design includes an emergency core cooling

system, it should be described and discussed in Chapter 6. The following auxiliary

systems that use primary coolant should be discussed in this section:

* experiment cooling

* experimental facility cooling

* experimental facility shielding (e.g., beam tubes)

* biological shield cooling

* thermal shield cooling

* fuel storage cooling and shielding

The applicant should provide the following information about these systems in this

section:

* design bases and functional requirements of the auxiliary systems based on

discussions elsewhere in the SAR, such as Chapters 4, 9, and 10,

'Experimental Facilities and Utilization"

* schematic drawings and flow diagrams that show the source of water,

locations of sensors and instruments, and locations of the components

cooled or shielded

* tables of the range of important parameters of the systems and

specifications of materials and components

* discussion of components to be cooled, the source of heat, the source of

the coolant water, heat transfer to the coolant, and coolant heat dissipation

* summary of the shielding requirements and the protection factors provided

by the coolant

* discussion of the provisions in the auxiliary system designs to prevent

interference with safe reactor shutdown

* discussion of the provisions in the auxiliary system design to prevent the

uncontrolled release of primary coolant or radiation exposures that would

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exceed the requirements in 10 CFR Part 20 and the facility ALARA

program guidelines

* requirements for Minimum water quality

* discussion of any technical specification requirements for the auxiliary

cooling systems, including the bases and surveillance requirements

5.8 Reference

Griess, J. C., et al., ORNL-3541, 'Effect of Heat Flux on the Corrosion of

Aluminum by Water," Part IV, Oak Ridge National Laboratory, Tennessee,

February 1964.

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6 ENGINEERED SAFETY FEATURES

In this chapter the applicant should discuss and descnie engineered safety features

(ESFs) for a non-power reactor. ESFs are active or passive features designed to

mitigate the consequences of accidents'and to keep radiologica exposures to the

public, the facility staf and the environment within acceptable values. The

concept of ESFs evolved from the defense-in-depti philosophy of multiple layers

of design features to prevent or mitigate the release of radioactive materials to the

environment during accident'conditions. .The need for'ESFs is determined by the

SAR analyses of accidents that could occur, even though prudent and conservative

design of the'facility has made the incidence of an accident very unlikely. It is also

possible that for a particular design the SAR analyses ivill show that ESFs are not

needed.

Normal operation of a non-power reactor is defined as operation with'all process

variables and other reactor parameters within allowed conditions of the license,

technical specifications, applicable regulatory limits, and design requirements for

the system. 'Accidents at non-power reactor facilities assume failure of a major

component such as the reactor coolant system boundary or a reactivity addition

event. Licensees analyze a maximum hypothetical accident that assumes an

incredible failure that leads to breach of the fuiel cladding or a fueled experiment

containment. These postulated accidents are compared with acceptance criteria

such as the safety limits from the technical specifications or, where there are

radiological consequences, to accepted regulatory limits (10 CFR Part 20 or 100).

The results'of the accident analyses are presented in SAR Chapter 13, *Accident

Analyses." ESF systems must be designed to function for the range of conditions

from normal operation through accident conditions.

Because most non-power reactors operate at atmospheric pressure, at relatively

low power levels, and with con e safety margins, few credible' postulated

accidents result in radiological risk to the public. The analyzed accident scenarios

that the applicant should present in SAR Chapter 13 include the following:

* loss of coolant

* ' loss of coolant flow '

* insertion of excess reactivity (rapid or ramp)

* loss of fuel cladding integrity or mishandling of fuel

* failure or malfunction of an experiment

* ' other uncontrolled release of radioactive material

* loss of electric power

*' external events such as floods and earthquakes

The SAR accident analyses for many non-power reactors may show that ESFs are

not required, even for the maximum hypothetical accident. In other cases the

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accident analyses may show that ESFs need to be considered in mitigating the

potential release of radioactive material to the environment.

The accident analyses provide the design bases for any required ESF. The ESF

design should be as basic and fail safe as practical. Because non-power reactors

are conservatively designed, few, if any, accidents should require redundant or

diverse ESF systems. However, consideration should be given to adding

redundancy and diversity to ESF systems if the reactor is of a higher power level

(2 MW or greater thermal power level), if an ESF system would be susceptible to

loss of capability to function because of a single failure, or if the radiological

consequences to the public of the accident that the ESF is designed to protect

against would be very serious if the ESF were to not function.

In addition to the design and functional characteristics of each ESF, the applicant

should describe the methods and criteria for testing to demonstrate ESF system

operability. The functional requirements, related setpoints, interlocks, and

bypasses for each ESF should be described, analyzed, and included in the facility

technical specifications. The technical specification surveillance requirements for

system components that ensure the integrity and operational capability of the ESFs

should be identified and discussed in the SAR

The discussion should include how the ESFs interact with site utilities, such as

electrical power and water and, if applicable, how the transfer between normal and

emergency sources of electricity and water is accomplished. The applicant should

discuss and demonstrate the need for site utility redundancy or diversity and the

specific design features that provide it for each ESF component.

The SAR should include schematic diagrams, showing all components, their

interrelationships, and the relationship of each ESF to other reactor systems (e.g.,

the core cooling system or the reactor room ventilation system). It should include

a brief description of the instrumentation and control (I&C) syste n for each ESF,

with detailed descriptions presented in SAR Chapter 7, 'Instrumentation and

Control Systems." The material presented should show how I&C systems

necessary for ESF operation are designed to function in the environment created

by the accident.

Typical ESFs that may be required at non-power reactors are (1) the confinement,

(2) the containment, and (3) the emergency core cooling system (ECCS). In

addition, features required in the facility heating, ventilation, and air conditioning

(HVAC) system to mitigate the consequences of accidents should be treated as

part of the ESFs of the confinement or containment system. The applicant should

discuss any additional ESFs in a comparable way.

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Brief definitions and illustrations of the confinement, the containnent, and the

ECCS follow:

(1) The confinement is an enclosure of the overall facility (e.g., a reactor

room) that is designed to limit the exchange of effluents between the

enclosure and its external environment to controlled or defined pathways.

A confinement should include the capability to maintain sufficient internal

negative pressure to ensure Wleakage (i.e., prevent uncontrolled leakage

outside the confined area), but need not be capable of supporting positive

internal pressure or significantly shielding the external environment from

internal sources of direct radiation.- Air movement in a confinement could

be integrated into the HVAC systems, including exhaust stacks or vents to

the external environment, filters, blowers, and dampers.

(2) The containment is an enclosure of the facility designed to (a) be at a

negative internal pressure to ensure inleakage, (b) control the release of

effluents to the environment, and (c) mitigate the consequences of certain

analyzed accidents. The containment is designed (a) to be sealed to

support a defined pressure differential across it and (b) to have a defined

upper limit on leakage from it. Both design conditions are testable. An

accident scenario that might require a containment for a non-power reactor

would involve positive internal pressures, either static or transient, or the

need to shield the external environment from internal sources of direct

radiation, or both. Exhaust stacks, vents, particulate filters, activated

charcoal filters, or piping may be provided for controlled venting of a

containment, and the design should provide for both normal 'nd emergency

operational modes. A containment may be designed to be integral with the

facility HVAC and liquid waste systems.

-Most non-power reactors can be designed, sited, and operated so that a

normal building or, at most, a confinement can be used to house the

reactor, a containment would not be required. In contrast to'a'

containment, a confinement plus its HVAC system

* usually responds to accidents by reducing and changing the airflow

paths to and from the building (a containment'seals the building.

from the environment and sigfiificantly reduces releases of

radioactive material to the environment)

* has doors with gasket-type seals (airlocks for contaimnents)

* nmay not have sealing isolation dampers on air penetrations (sealing

isolation dampers for containments)

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* cannot maintain as high a negative differential pressure as a

containment

* is not as leaktight as a containment and the leak rate normally

cannot be confirmed through testing

* cannot control the release from an event that results in positive

pressure in the reactor building

* usually has less direct radiation shielding capacity than a

containment because the walls are thinner

* is less resistant than a containment to challenges placed on the

building by the external environment

If the analyses show that a confinement ESF will mitigate the consequences

of the most limiting accident scenario to acceptable levels, a containment

ESF would not be required, although some licensees have chosen to build

containments as an additional design conservatism.

(3) An ECCS is designed to provide a source of coolant to limit fuel damage

from decay heat should primary cooling be lost from the reactor core

region.

The issue of what standards to use in evaluating accidents at a non-power reactor

was discussed in an Atomic Safety and Licensing Appeal Board (ASLAB) decision

issued May 18, 1972, for the research reactor at Columbia University in New York

City. The ASLAB stated that "as a general proposition, the Appeal Board does

not consider it desirable to use the standards of 10 CFR Part 20 for evaluating the

effects of a postulated accident in a research reactor inasmuch as they are unduly

restrictive for that purpose. The Appeal Board strongly recommends that specific

standards for the evaluation of an accident situation in a research reactor be

formulated." The NRC staff has not found it necessary to follow the board

recommendation to develop separate criteria for the evaluation of research reactor

accidents, since most research reactors to date have been able to meet the

conservative 10 CFR Part 20 criteria. American National Standards

Institute/American Nuclear Society ANSIIANS-15.7, "Research Reactor Site

Evaluation," contains additional information on doses to the public from releases

of radioactive material.

The objective of non-power reactor ESFs is to ensure that projected radiological

exposures from accidents are kept below the regulatory limits. The regulations

defining the limits on releases from non-power reactors during accident conditions

depend on whether the non-power reactor (see 10 CFR 50.2) is a test reactor (also

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ENmINERED SAFETY FEATURES

called a testing fcility, see 10 CFR 50.2) or a research reactor (see 10 CFR

170.3). For a research reactor, the results of the accident analyses have generally

been limited to values from the old 10 CFRPart 20 (10 CFR 20.1 through 20.602

and appendices). The exposure values that the staff has generally found acceptable

for research reactors are less than 5 rem whole body and 30 rem thyroid for

occuationally exposed persons and less than 0.5 rem whole body and 3 rem

thyroid for members of the public. However, in several instances, the staff has

accepted very conservative accident analyses with results greater than the 10 CFR

Part 20 dose limits discussed above. Research reactors that received fthir initial

operating license after January 1, 1994, must show that exposures meet the -

requir ts of the revised 10 CFR Part 20 (10 CFR 20.1001 through 20.2402

and appendixes). Occupational exposure is discussed in 10 CFR 20.1201, and

public exposure is discussed in 10 CFR 20.1301.

If the facility meets the definition of a test reactor, the exposures will be compared

with the doses in 10 CFR Part 100. As discussed in the footnotes to 0 CFR -

100.11, the doses in 10 CFRPart 100 are reference values. References to 10 CFR

Part 100 in this chapter pertain to test reactors only.

6.1 Summary Description

In this section of the SAR, the applicant should briefly describe all of the ESFs in

the facility design and summariz= the postulated accidents they are designed to

mitigate. These summaries should include the design bases and performance

criteria and contain enough information for an overall understanding of the

functions of the ESFs and the reactor conditions under which the equipment or

systems must fimction.

Simple block diagrams and drawings may be used to show the location, basic

function, and relationship of each ESF to the facility. Detailed drawings, -

schematic diagrams, data, and analyses should be presented in subsequent sections

of this chapter for specific ESFs.

6.2 Detailed Descriptions

In this section of the SAR, the applicant should discuss in detail the particular

ESFs incorporated into the reactor desig Not all of these ESFs are found in any

single design. Other systems in addition to the systems discussed in this section

may be considered ESFs The applicant should discuss these ESFs in a manner

similar to the discussions in this section.-;

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6.2.1 Confinement

The applicant should discuss in detail the confinement and the associated HVAC

systems that function as ESFs. For the confinement to function as an ESF, the

design bases for the consequence-mitigation functions should be derived from the

accident analyses in SAR Chapter 13.

Confinements and HVAC systems may also have functions that are not considered

functions of ESFs and that need not be addressed in this chapter. Most non-power

reactors release small quantities of airborne radioactive material, primarily argon41 gas, to the environment during normal operations. To protect the health and

safety of the public and the staff, it may be necessary to control airflow through the

reactor room and release the reactor room air in a controlled manner at a location

that allows for dilution and diffusion of the radioactive material before it comes in

contact with the public. In some cases, it may also be efficient to use the

confinement and HVAC systems to prevent an uncontrolled release to the

environment of radioactive effluents resulting from operation. This aspect of the

use of ESFs during normal system operation is not considered an ESF function.

However, the design bases and detailed discussions of these systems for normal

operations to control releases should be given in Chapter 3, "Design of Structures,

Systems, and Components," and Chapter 9, "Auxiliary Systems." Diffusion and

dispersion of airborne radioactivity in both restricted and unrestricted

environments should be discussed in Chapter 11, "Radiation Protection Program

and Waste Management."

A radioactive release need not be a rapid or burst-type release. It also includes the

leakage and diffusion of airborne radioactivity from a room through cracks or gaps

in building structural components. Such releases could be controlled by a system

of ducts, louvers, blowers, exhaust vents, or stacks. Non-power reactors should

have the capability to quantify releases and calculate potential exposures in both

restricted and unrestricted areas. Calculating potential exposures provides the

bases for actions to ensure that the public is protected during both normal

operation and accident conditions.

If the confinement and HVAC or air (stack) exhaust systems are designed to

change state or operating condition in response to a potential accident and, in so

doing, mitigate the radiological consequences of the accident, those features

should be designated as ESFs and should be described in detail. The discussion of

the ESF functions should demonstrate how dispersion or distribution of

contaminated air to the environment or occupied spaces other than the reactor

room is controlled. The discussion should include the design bases for the location

and operating characteristics of the air exhaust stack, if applicable, and the design

bases for effluent monitoring systems.

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The discussion of mitigative effects should contain a comparison of potential

radiological exposures to the facility staff and the public with and without the ESF.

Either operational data for an operating facility or results of nalyses for a new

faciity should be presented showing airflow rates, reduction in quantities of

airborne radioactive material by filter systems, system isolation, and other

parameters that demonstrate the effectiveness of the system.

A schematic diagram of the system should be presented showing the blowers,

dampers, filters, other components necessary for operation of the system and flow

paths. Automatic and manual trip circuits, bypasses, interlocks, and special I&C

systems for the ESF system should be described briefly in this 'section' and in detail

in Chapter 7.

In this section the applicant should develop requirements to be specified in the

technical specifications for system operability, periodic surveillance, setpoints, and

other specific requirements to ensure a functional ESF system during postulated

events. Examples include the requirement for operability of the ESFs during

reactor operation or other significant events such as fuel movement. Periodic'

functional testing of damper closure, room isolation, minimum airflow rates,

automatic system shutdown and startup, and activation setpoints should be

required and specified. See Chapter 14, 'Technical Specifications," of this format

and content guide, for details on what technical specification requirements should

be identified and justified in this section.

6.2.2 Containment

Because the potential risk to the public is generally low from accidents at nonpower reactors, few require a containment. However, a containment and

associated HVAC system required at a non-power reactor to mitigate the ?

consequences of a postulated accident are considered ESFs. As previously

discussed, containments are required as an ESF on the basis of the reactor facility

design, operating characteristics, accident scenarios, and location.

A containment for a non-power reactor should be designed to'prevent 'the rapid,

uncontrolled release of radioactive material to the environment. A possible

scenario for such a release could be an accident in the reactor core that involves a

large loss of fuel cladding integrity (multiple plates or pins),'the release of fission

products into the primary coolant system, and the rapid release of fission products

from the primary coolant system into the reactor room.' The containment is

designed to control the release to the'environment of airborne radioactive material

released in the reactor room even if the'accident is accompanied by a pressure

surge 'or a steam release within'the room. The thick walls of the containment may

also help mitigate direct radiation exposure during certain accidents. The analyses

in Chapter 13 of the SAR should include details of the postulated scenario,

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including the assumptions and justification for the initiating event, the progression

of the scenario, the consequence-mitigating effects of the containment, and the

potential radiological exposures to the most exposed member of the public. The

design bases for the containment should include the postulated peak pressures, the

duration of the event, the pressure-versus-time envelope, the time during which

containment integrity must be maintained while recovery from the event is

implemented, limits on leakage or controlled release from the containment to the

environment, the quantity of failed fuel, and the quantity and type of released

radioactive material.

A radioactive release need not be a rapid or burst-type release. It also includes the

leakage and diffusion of airborne radioactivity from a room through cracks or gaps

in building structural components. Such releases could be controlled by a system

of ducts, louvers, blowers, exhaust vents, or stacks. Non-power reactors should

have the capability to quantify releases and calculate potential exposures in both

restricted and unrestricted areas. Calculating potential exposures provides the

bases for actions to ensure that the public is protected during both normal

operation and accident conditions.

The description must include the bases for the protection factors provided by the

containment. The goal is that the containment should reduce the consequences to

the public, the facility personnel, and the environment to acceptable values as

specified above.

In this section the applicant should explain how the design and functional details of

the containment meet the design bases and criteria described above. System

drawings, component and material specifications, and structural details should be

included. The information should demonstrate that the radiation protection factors

assumed in the accident analyses are provided. The design bases and the

discussions should describe how the containment functions over the range of

normal operation and the events that initiate switching to the emergency mode.

The discussions should address which reactor operations and evolutions require

the containment to be operable, and whether an emergency electrical power source

is required to be operable.

To qualify as a containment, the reactor building should be a robust structure with

airlocks and all other penetrations sealed (e.g., cable penetrations sealed with

epoxy) or sealable (e.g., hydraulic dampers on ventilation penetrations). The

building should be capable of maintaining a negative pressure in relation to the

atmosphere (e.g., at least -1/2 inch water) during normal operation and have a

measurable leakage rate (e.g., less than 5 percent over 24 hours). The actual

performance requirements are determined from the accident analyses in Chapter 13

of the SAR. For example, the normal function of the containment ventilation

exhaust system may be divided into two trains-one that ventilates the reactor

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ENGIEERE SAFETY FEATUREs

room and a second that ventilates areas with high airborne radiation generation

such as experimental ficilities or fume hoods. The ventilation system is normally

equipped with high-efficiency particulate filters, and the accident ventilation system

has a separate train(s) equipped with high-efficiency particulate and activated:

charcoal filters to sorb iodine.

Automatic containment trip circuits, interlocks, special I&Cs, and monitoring

requments for the ESF should be described. The description should detail their

relationship and interaction with the I&C systems for normal operation described

in SAR Chapter 7.

The discussion should give the technical specifications and their bases to ensure

that the containment ESF is operable when required. The technical spmcifications

also should provide for necessary surveillance, testing, and maintenance of the

containment components to ensure operability. The technical specifications should

define an operable containment ESF and describe the reactor conditions and

operations for which the containment shall be operable. See Chapter 14 of this

format and content guide for details on what technical specification requirements

should be identified and justified in this section.

6.2.3 Emergency Core Cooling System

An ECCS may be required at some non-power reactors to remove decay heat from

the fuel to prevent fiure or degradation of the cladding if cooling is lost. The

applicant should give the analysis of the ECCS if one was identified as needed in

the Chapter 13 accident analyses.

A schematic diagram should show the relationships among the major system

components such as valves, spray headers, pumps, piping, and any I&C systems.

Special ECCS I&C systems should be described briefly in this section and

described fully in Chapter 7. In this section, the applicant should discuss any

effects of the ECCS design on normal operations and reactor safety. Analyses for

non-power reactors should demonstrate that fuel failure will be prevented for

postulated-accident scenarios.

If the ECCS is a passive system (e.g., a gravity-feed spray from a storage tank), a

complete description with associated analyses and data should show how coolant

flow is initiated and why the system is effective. The information should

demonstrate that the ECCS will provide core cooling capacity in terms of

minimum flow and time of operation for all loss-of-coolant accidents considered.

If the ECCS is an active system that requires sensors and an action or event to

initiate operation, descriptions should include details of initiation response times

and backup or redundant sensing and control systems. The discussion should

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include the source of electrical power, source of coolant, heat sink, or other

systems required to operate the ECCS and show how operability and availability

are ensured.

The facility design should show how radioactive materiaL such as emergency

coolant, is controlled.

In this section the applicant should also give the bases for technical specifications

that ensure that the ECCS is available and operable when required. Technical

specifications should include minimumn operability requirements and the possible

operations and conditions under which the ECCS would be required. Test and

surveillance functions and intervals should be stated in the technical specifications

to ensure operability of the ECCS. See Chapter 14 of this format and content

guide for details on what technical specification requirements should be identified

and analyzed in this section.

6.3 References

American National Standards Institute/American Nuclear Society, ANSI/ANS15.7, 'Research Reactor Site Evaluation," ANS, LaGrange Park, Illinois, 1977.

Atornic Safety and Licensing Appeal Board, 'In the Matter of Trustees of

Columbia University in the City of New York," May 18, 1972.

NIJREG-1537,PART I 6-10 REV 0.2/96

NUREG-1537,PART I 6-10 REV 0, 2/96
7 INSTRUMENTATION AND CONTROL SYSTEMS

In this chapter of the SAR the applicant describes and discusses the design and

operating characteristics of the instrumentation and control (I&C) systems.

Sufficient information should be included to explain the design criteria and bases,

and to discuss the functional and safety analyses of the I&C subsystems. The I&C

subsystems generally comprise the reactor control system (RCS), process

instrnuments, the reactor protection (safety) systemn(RPS), instruments to initiate

operation of engineered safety features (ESFs), and radiation safety monitoring

systems. These systems and their outputs can be consolidated into a control

console, along with the devices and circuits that control the operation of the

reactor. The guidance in this chapter of the SAR is based on the principle that

most non-power reactors can be designed and operated to pose acceptably small or

insignificant risk to the public without isolating or separating the RPS from other

subsystems. Additional design features, such as separation' of systems, may be

necessary for high-power test reactors. 'Applicants who need additional guidance

beyond that given in this chapter should contact their project manager.

The non-power RPS should monitor selected reactor'operating parameters such as

neutron flux; fuel temperature (monitored primarily in TRIGA-type reactors);

primary coolant flow, temperature, and level; and radiation intensity. The RPS is

designed to ensure reactor and personnel safety by limiting parameters to operate

within analyzed operating ranges. The RPS can also give the ESF actuation

system information for the operation ofESFs when the instruments indicate that

abnormal or accident conditions could occur. The RCS may monitor many of the

same parameters as the RPS and give information for automatic or manual control

of the reactor operating conditions (e.g., reactor power, by inserting or

withdrawing control rods). The reactor facility instruments present operating

parameter and system status information to the operator for monitoring reactor

operation and for deciding on manual control actions to be taken. Instrument

systems are the means through which automatic'or operator control actions are

transmitted for execution by the RCS. Radiation instruments 'show radiation levels

in selected areas'in'theireactorifacility and could give data to the RPS, 'give

information to help in the control of personnel radiation exposure, 'or monitor the

release of radioactive material from the reactor and the reactor building.'

In this chapter, the applicant should discuss the fiiuntional requirements, design

criteria and bases, system descriptions, system performance analyses, and the bases

of technical specification limiting safety system settings (LSSSs), lmiting

conditions of operation (LCOs), and surveillance requirements for the I&C

systems for non-power reactors.

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7.1 Summary Description

In this section of the SAR, the applicant should briefly describe the I&C systems of

the reactor, including block, logic, and flow diagrams showing major components

and subsystems, and connections among them. The applicant should summarize

the technical aspects, safety, philosophy, and objectives of the I&C system design

and should discuss such factors as redundancy, diversity, and isolation of

functions. The information should include:

Type of instruments-System instruments should be described by type

[e.g., hardwired analog, computerized digital that uses stored programs

(software), or combinations of these]. If a combination is used, the

applicant should clearly note which portions or functions are analog and

which are computerized digital, and how they relate to each other. The

applicant could refer to existing systems reviewed and approved by NRC

that are similar to the described system.

* Classification of systems-I&C systems and equipment should be classified

into categories by function performed (e.g., the RCS, RPS, ESF actuation

system, control console and display instrument systems, and radiation

protection instruments).

The general description of each category of I&C subsystem should include the

following, as applicable:

* For the RCS, a brief discussion of each major subsystem such as manual

control system, automatic control system, control rod drive systems,

bypass and interlock systems, and any integrated experiment I&C systems.

* For the RPS, the types of parameters monitored, both nuclear and nonnuclear, the number of channels designed to monitor each parameter, the

actuating logic that determines the need for actions to change reactor

conditions and that takes these actions, and number and type of reactivity

control devices.

* For the ESF actuation system, a discussion of the subsystems that detect

the need for operation and that initiate operation including identification of

the parameters monitored or the source of input information and the

number of channels designed to monitor, process, and act on the

information.

* For the control console and display instruments, a discussion of the

parameter display systems and equipment by which the operator can

observe and control the operation of the reactor and important subsystems

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For radiation protection instruments, a brief discussion of area and effluent

radiation detection systems that monitor, alarm, or provide input to other

subsystems of potentially hazaidous radiation levels. The applicant should

address radiation system that monitor effluent streams from the reactor

-facility, state the type of effluent (such as airborne or liquid), and list aarms

or signals to other subsystems.

* A sunmary of the human-machine interface principles used in the location

of instrumentation and controls.

7.2 Design of Instrumentation and Control Systems

7.2.1 Design Criteria

In this section of the SAR, the applicant should discuss the criteria for developing

the design bases for the I&C systems. The basis for evaluating the reliability and

performance of the I&C systems should be included. All systems and components

of the I&C systems should be designed, constructed, and tested to quality

standards commensurate with the safety importance of the functions to be

performed. Where geneially recognized codes and standards are used, they should

be named and evaluated for applicability, adequacy, and sufficientcy. They should

be supplemented or modified as needed in keeping with the safety importance of

the function to be performed.' Evaluations and modifications of the standards

should be described in the SAR. A set of generally applicable criteria for use as a

guide is given below. Criteria that are used should be clearly stated and should be

shown to provide the appropriate level of reliability, safety, and performance

capability. The applicability of these criteria should be determined from the

operating analyses in Chapter 4, 'Reactor Description," and accident analyses in

Chapter 13, "Accident Analyses," of the SAR'.

* Systems and compoinients (including I&C systems) determined by the

analyses in'the SAR to'be important to the safe operation or shutdown of

the reactor should be designed to be in accordance with local building and

siting c6des, and should beable tio withstand the effects of natual - phenomena without loss of capability to performrtheir safety function (see

Chapter 3, 'Design of Structures, Systems, and Components," for

additional information).

* Ic&C systems and components determined in the SAR-analyses to be

important to the safe operation or shutdown of the reactor should be

'designed, located, and uirottectid so thdatthe effects of fires or explosions

would not prevent them from performing their safety finctions.

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I&C systems and components determined in the SAR to be important to

the safe operation or shutdown of the reactor should be designed to

function reliably under anticipated environmental conditions (e.g.,

temperature, pressure, humidity, and corrosive atmospheres) for the full

range of reactor operation, during maintenance, while testing, and under

postulated accident conditions, if the systems and components are assumed

to function in the accident analysis.

* The RPS should be designed to automatically initiate the operation of

systems or give clear warning to the operator to ensure that specified

reactor design limits are not exceeded as a result of measured parameters

indicating the onset of potential abnormal conditions. The ESF actuation

system should be designed to automatically initiate operation to mitigate

the consequences of abnormal conditions or accidents.

* I&C systems should be designed to have functional reliability, including

redundancy and diversity, commensurate with the safety functions to be

performed and the consequences of failure of the system to perform the

safety function. For example, an I&C system for a non-power reactor

should be designed to perform its protective function after experiencing a

single random active failure within the system.

* I&C systems should be designed to fail into a safe state on loss of electrical

power or exposure to extreme adverse environments.

* I&C systems should be designed so that a single failure will not prevent the

safe shutdown of the reactor.

7.2.2 Design-Basis Requirements

I&C system design requirements for non-power reactors are generally derived

from the results of analyses of normal operating conditions and of accidents and

transients that could occur. This section provides guidance on the factors to

consider in developing the analyses and the design bases. Design bases for the

I&C system, subsystems, and components should include the following, as

applicable:

* The function or purpose of systems or instruments considering which

reactor parameters are monitored or controlled.

* The range of values that monitored variables may exhibit for normal

operation, shutdown conditions, and for postulated accidents.

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* Safety or control functions and any unique or facility-specific functions

performed by the I&C system or subsystems.

* 'Specification of alarm, trip,'and actuation setpoints derived from accident

or other operational analyses of the instrumented systean or function.

* Any special requirements such as redundancy, diversity, quality assurance,

and environmental requirements 'derived from the results of analyses of the

full range of operating conditions and postulated accidents.

* 'The specification of precision and accuracy requirements for the

instruments, control subsystems, or components.

* The specification of number and type of channels required to monitor

variables.

* The system operational and support requirements such as those for

electrical, mechanical, structural, cooling, heating, and signal input.

* The requirements that controls and instruments be grouped and located so

that operators can easily reach and manipulate the controls while readily

observing on meters and displays the results of their actions (operator

interface requirements).

* For digital computer systems, in addition to the foregoing, the applicable

guidelines from IEEE 7-4.3.2-1993, JEEE Standard Criteria for Digital

Computers Systems in Safety Systems of Nuclear Power Generating

Stations," for the design, application,-and evaluation of digital computer

hardware and software and ANSIIANS-10.4-1987," Guidelines for the'

Verification and Validation of Scientific and Engineering Computer

Programs for the Nuclear Industry," for evaluating the verification and

validation programs for software for use in the I&C system. Regulatory

Guide (RG) 1.152, which discusses the use of digital computers in nuclear

power plant safety systems is attached as Appendix 7.1. However, neither

of these standards was uniquely developed for non-power reactors and may

contain sections and requirements that do not apply to a particular,

situation. Furthermore, the technology and safety principles on which

computerized I&C systems are based are changing. Until additional

'guidance is available, applicants should request current requirements from

-NRC. The testing programs for compputer systems help to verify that the

software will not cause unintended effects under some combinations of

circumstances or conditions, or some malfunctions. ANSVANS Standard

15.15-1978, Criteria for the Reactor Safety Systerns of Research

Reactors," and draft ANSI/ANS Standard 15.20, "Criteria for the Control

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CHAPTER 7

and Safety Systems for Research Reactors," are general guides for the

design, implementation, and evaluation of I&C systems for non-power

reactors and should be used where applicable. The staff reviewed and

accepted a digital control system developed by General Atomics. NRC

licensees for several TRIGA reactors have installed this system (see

Amendment No. 19 to Facility Operating License No. R-84, Docket No.

50-170 for the Armed Forces Radiobiology Research Institute TRIGA

reactor, July 23, 1990. Amendment No. 29 to Facility Operating License

No. R-38 for the General Atomics TRIGA Mark I reactor is attached to

this chapter as Appendix 7.2). The staff reviewed and accepted a digital

control system developed by Atomic Energy Canada Limited. An NRC

licensee for a TRIGA reactor installed this system (Amendment No. 30 to

Facility Operating License No. R-2 for the Penn State Breazeale Reactor is

attached to this chapter as Appendix 7.3)

Consult NRC Generic Letter 95-02 for I&C systems that are being

upgraded to systems with digital technology; it is attached to this chapter

as Appendix 7.4.

7.2.3 System Description

The system description in the SAR should include equipment and major

components as well as block, logic, and schematic diagrams. The applicant should

also submit hardware and software descriptions and software flow diagrams for

digital computer systems. The applicant should describe how the system

operational and support requirements will be met and how the operator interface

requirements will be met. The description should also address the methodology

and acceptance criteria used to establish and calibrate the trip or actuation

setpoints, or interlock functions.

7.2.4 System Performance Analysis

The applicant should conduct a performance analysis of the proposed I&C system

to ensure the design criteria and design bases are met and license requirements for

the performance of the system are specified. The system performance analysis

should encompass the following:

* The SAR should describe the operation of the I&C system and present the

analysis of how the system design meets the design criteria and design

bases. The discussion should include accuracy, reliability, adequacy and

timeliness of I&C system action, trip setpoint drif, quality of components

and, if required by the analyses, redundancy, independence, and how a

single failure affects both its ability to perform its safety function and the

effect on operation or safe shutdown of the reactor.

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* Technical specification LSSSs, LCOs, and surveillance requirements for the

I&C system should be established. These parameters and requirements

should include system opeaility tests, trip or actuation setpoint checks,

trip or actuation-setpoint calibrations, and any system response-time tests

that are required. Surveillance inteivals should be specified and the bases

for the intervals, including operating experience, engineering judgment, or

vendor recommendation. should be discussed.

7.2.5 Conclusion.

The applicant should summarize in this section of the SAR why the system design

is sufficient and suitable for performing the functions stated in the design bases.

7.3 Reactor Control System.

The RCS performs several functions, such as maintaining the reactor in a

shutdown state, reactor startup, changing power levels, maintaining operation at a

set power level, and shutting down the reactor. In non-power reactor designs that

allow pulsing such as the TRIGA design); the RCS can rapidly insert reactivity into

the reactor core to produce a predetermined high-power pulse of short duration, or

to achieve a rapid increase in reactor power in a "square wave." The RCS may be

discussed using such subsystems as nuclear instruments, process instruments,

control elements, and interlocks. In describing each subsystem in the SAR, the

applicant should include design considerations and technical specification

requirements.

In the nuclear instiument system, nuclear instruments monitor the neutron flux'

from the subcritical source multiplication range, through the critical range, and'

through the intermediate flux range to full power. Neutron flux i=struments also

should determine the startup rate and, im some designs,reactor period information.

Linear and log neutron flux channels should be used to monitor the core neutron

flux while control rods are withdrawn or inserted to increase or decrease reactor

-power. At least one linear neutron flux channel should be calibrated to reactor

thermal power.

The process instruments are designed to measure and display such parameters as

coolant flow, temperature, or level; fuel temperature; 'or'air flow parameters within

or from the reactor room. In some designs, this information may also be sent to

the RPS.

The typical non-power reactor has an automatic control (servo) system that

controls the reactor power about a point set by the operator. Most servo control

systems compare the output of a linear neutron flux channel against an adjustable

voltage representing the desired power level; and automatically change the position

REv. 0,2/96 1-7

REV. 0, 2/96 .- .7-7
CHAPTER 7

of a regulating rod in the core to change the neutron flux density to reduce the

difference between the two voltages until the actual reactor power level is very

nearly equal to the desired power level. This process can be performed by analog

control equipment or by software in a digital computer system.

Reactors with pulsing capabilities have a transient rod that, on command, is rapidly

ejected out of the core to a pre-programmned distance. This action rapidly inserts a

known amount of excess reactivity into the core that pulses the core power to very

high levels for very short intervals. The system can also be used to form a squarewave power increase to a predetermined steady-state power level.

The RCS for non-power reactors should have a set of equipment protection

interlocks and inhibits that prohibit or restrict operation of the reactor unless

certain conditions are met. For example, there should be an interlock that

prohibits control rod motion unless the neutron flux in the core produces a neutron

count rate sufficient to help ensure that nuclear instruments are responding to

neutrons. There may be additional equipment protection interlocks to ensure, for

example, that there is sufficient coolant flow, shielding is intact, ventilation air is

flowing, coolant level is sufficient, and required neutron instruments and recorders

are functional. There may also be personnel protection interlocks to prevent

reactor operation if certain radiation fields are excessive. Control rods may be run

back to automatically reduce the reactor power when certain specified reactor

conditions approach a predetermined limit, but total reactor shutdown (scram) is

not warranted.

Experimental facilities may be interlocked with the RCS to prevent reactor

operation if the experimental facility is not in the correct configuration. If

experiments conducted in non-power reactors could interact with the core to

change reactivity or otherwise modify the reactor operating conditions, data to the

RCS or RPS from the experiment instruments may be needed to detect reactivity

changes. All experiments should be carefilly considered for interaction with the

I&C system when the safety analysis for the experiment is performed. The analysis

should consider any interaction with the RCS or RPS. Where such interactions are

warranted, they should meet the standards used for the design of the systems to

which the experimental facilities will be connected.

The applicant should include the following for each RCS subsystem:

* Discuss the design criteria for the RCS as outlined in Section 7.2.1,

including any criteria specific to the reactor design not outlined in the

section.

* Discuss the design bases information specified in Section 7.2.2 and any

additional design bases of facility-specific subsystems.

NUREG-1537, PART I 7-8 REV.O,2/96

NUREG- 1537, PART I 7-8 REV. O. 2196
INSTRUMENTATION AND CONTROL SYSTEMS

* Describe the system as specified in Section 7.2.3, including any additional

system descriptive material specific to subsystem design and

implementation not covered in Section 7.2.

* Analyze the operation and peiformane of the system as specified in

Section 7.2.4 including analyses and results of any features or aspects

specific to the facility design and implementation not specified in

Section 7.2. Include the bases of any technical specifications and

surveillance tests with intervals specific to the design and operation of the

systems. Address the specific design features of the RCS, such as the

following:

Detector channels directly monitor the neutron flux density for

presentation of reactor power level and power rate-of-change.

The RCS gives a continuous indication of the neutron flux density

from subcritical multiplication source level through the full licensed

power range. If multiple detector channels are used, this

continuous indication should overlap a minimum of one decade

during detector changeover.

The RCS has a reactor period channel that covers subcritical

neutron source multiplication from the approach to critical, through

critical, and into the licensed power range. Depending on the

analysis in the SAR, some reactors may not have this channel.

The RCS protects against a failure or operation in a mode that

could prevent the RPS from performing its intended safety function.

- The system and equipment are designed to assume a safe state on

loss of electrical power.

- The RCS has at least two channels of reactor power indication

through the licensed power range.

- The startup and operating power detector channels can discriminate

against strong gamma radiation, such as that present after long

periods of operation at full power, to ensure that indicated changes

in neutron flux density are reliable.

- The reactor power indication of at least one channel should remain

reliable for some predetermined range above the licensed power

level. For reactors with power level as a safety lirmit, the

instrumentation should be able to indicate if the safety limit was

exceeded. For other reactor types, at least one channel should be

RE'. 0.2196 7.9 STANDARD FORMAT AND CONTENT

REV. O. 2/96 7-9 STANDARD FoRmAT AND CoNTFmr
CHAPTER7

able to indicate if the power level was exceeded, which is the basis

for limiting licensed power leveL

- All control rod positions should be indicated at the control console

throughout their travel and should indicate when they are at an "in"

or "out" limit.

- The control rod drive speed in "manual" and "automatic" modes of

operation should be limited to that analyzed and allowed for

controlling the rate of change of reactivity.

- While in 'automatic" reactor control mode, the RCS should

indicate being placed in or removed from automatic control.

- In a reactor designed for pulsing, the movement of the transient rod

should be limited in accordance with reactivity amounts and rates

derived from the SAR analysis.

- In a reactor designed for pulsing, the system should indicate the

position of the transient rod, when this rod is fully inserted, and

when set in position to initiate a pulse, and should provide

interlocks to ensure the position of the rod.

- Bypasses of interlocks should be under the direct control of the

reactor operator and should be indicated in the control room.

- The RCS and the reactor reactivity control system should meet the

requirements of minimum shutdown margin considering the stuck

rod criteria

The applicant should discuss the conclusions about capability and suitability

of the RCS requested in Section 7.2.5.

7.4 Reactor Protection System

The RPS is designed to detect the need to place the reactor in a subcritical, safe

shutdown condition (scram) when any of the monitored parametIers exceeds the

limit as determined in the SAR Upon detecting the need, the RPS should

promptly and automatically place the reactor in a subcritical, safe-shutdown

condition (scram) and maintain it there. This prevents or mitigates unintended

operation in regions where risks of the following types could occur: fuel damage

from overpow er or loss of cooling events, uncontrolled' release of radioactive

materials to the unrestricted environment, or overexposure of personnel to

radiation. Parameters monitored for this purpose could include core neutron flux,

NUREG-1537, PART I 7-10 REv 0.2/96

NUREG-1537,PART I 7-10 REV 0, 2/96
Ix)UMDrrAIIoN AND CoNotO SysIrhms

fuel temperature, core coolant flow and temperature, coolant level, area radiation

levels, and air concentration, or release, of radioactive materials.

Non-power reactors can be designed and operated so that postulated accidents

pose risks to the facility or the public that are not significant or that are within

applicable regulatory limits. Ifjustified by the accident analyses of Chapter 13, the

RPS need not be separate and independent of the RCS. The applicant for such

reactors may perform an analysis to determine whether certain RPS-monitored

parameters or interlocks should be required to be redundant, diverse, or singlefailhire-proot Two'examples of these parameters are the reactor pool level or area

*'radiation exposure rates. Therefore, the RPS and its subsystems should be

designed in accordance with the guidance in Section 7.2, and the'SAR should

include the following information:

* Discussion of design criteria for the RPS as outlined in Section 7.2.1,

including any criteria specific to the reactor design not outlined in the

section.

* Safety and system design bases information'as specified i Section 7.2.2.

Any supplemental facility-specific design bases not specified in the general

system requirements should be included.

* System descriptions consistent with that specified in Section 7.2.3, along

with any subsystem description that is ficility specific and that may not be

identified in the general system reqiments.

* Analyses of the operation and performance of the RPS similar to that

specified in Section 7.2.4. This should include analysis of any features,

aspects, or technicl specifications including surveillance tests that may be

reactor specific and not identified in'the'general'system requirements.

These analyses should be based on postulated credible accidents, transients,

and other events that could require RPS intervention, and should include all

of the applicable features noted'in Section 7.3 fo'r'the RCS. The analyses

should include quantitative performance of all scrams, runbacks, interlocks,

and ESF initiators. The specific design features of the RPS that should be

addressed indlude'the'following: -

- Independent redundant or diverse reactor power level trips.

- A log power level channel with a reactor period or rate-of-flux

change output with a rate or period channel set to scram in

accordance with the analysis (certain reactor designs' do not require

the period scram design feature because they are designed to

accommodate rapid additions of reactivity). The log channel and a

RE.029 -1SADR OMTADC-~

REV. 0,2/96 . . - 7-11 -- - STAmmDFoRmATAmCoNTENT
CHAPTE7 -

linear flux monitoring channel should accurately sense neutrons

even in the presence of intense high gamma radiation.

Neutron flux (power) monitor channels covering the range from

subcritical source multiplication to well beyond the licensed

maximum power level.

A startup channel measuring neutrons at subcritical with a minimum

count rate interlock to ensure operation and to' prevent control or

safety rod withdrawal unless the neutron count rate is at least some

predetermined minimum such as 2 counts per second. This

interlock may not be needed in reactor designs that use

photoneutrons for startup. The applicant should justiiy not needing

the interlock in this case. The detector is capable of detecting

neutrons in a high gamma field and can be verified so that

subcritical neutron multiplication can be determined and all

reactivity changes can be monitored until the startup channel

indication is overlapped by the log or linear channel power

indication.

RPS scram time as established in the accident analysis, and any

other requirements to ensure operability.

The scram circuit is designed for the protective action to go to

completion once it is initiated. The circuit cannot be reset until all

released rods are filly inserted.

Each scram channel has a separate set of contacts or other bi-stable

component to trip the RPS system.

The manual scram switch is located where the operator has ready

access, such' as near the rod drive controls.

- Upon receipt of a scram signal, the RPS will annunciate the scram

and signify the circuits that are in a tripped state.

- The RPS shall always be capable of shutting down the reactor at

least to the shutdown margin defined in the technical specifications.

Conclusions about capability, operability, and suitability of the RPS

requested in Section 7.2.5.

_UE-57 ATI71 E.029

NUREG- 1537, PART I 7-12 REV. 0, 2/96
INsTumE rAI1ON AND CONTROL SYSTEMS

7.5 Engineered Safety Features Actuation Systems

If ESFs are required by the accident analyses in Chapter 13, their actuation

systems should be described in this section. The ESF actuation system'senses the

need for and initiates the operation of ESF systems (1) to prevent or mitigate the

consequences of damage to fission product barriers such as fuel, claddifig, or

fueled experiments caused by overpower or loss-of-cooling events or (2) to gain

control of any radioactive material released by accidents.

Each active ESF should be automatically initiated by a'subsy'stem of the ESF

actuation system. Examples of such systems include those to actuate an active

emergency core cooling system (ECCS), containment or confinement system,

containment or confinement air cleanup 'and filtration system, or any other ESF

that is'designed to perform a mitigative function. Most non-power reactors do not

have an active ECCS because they are'designed to'rely on passive ECCS or natural

coolant circulation to provide sufficient core cooling to prevent loss of fuel

integrity. Certain non-power reactors'nay fnot be'required by the accident analyses

to have containment or confinement ESF systems ora containment or confinement

air cleanup and filtration ESF system. When such systems'are required, their

actuation systems should be described in this section, in coordination with the

information in Chapter 6, "Engineered Safety Features," of the SARCertain parameters should be monitored to determine the need to initiate the

operation ofESFs.' These parameters should be determined by the accident

analyses, and may include fuiel temperature, core coolant flow and temperature,

coolant level, area radiation, and radioactivity of airborne materials. ESF actuation

systems need not be designed to be redundant or diverse, orto be able to survive a

single failure and still perform the safety function unless the accident analysis

requires these features.

The applicant should describe the ESF actuation system in sufficient detail to

describe the functions required of the ESF and the operation of the syte. The

SAR should include the following information for each required ESF actuation

system:..

* Design criteria for the ESF actuation system as outlined in Section 7.2.1,

including any criteria specific to the reactor design not outlined in the

section..

* Design bases information for the ESF actuation system as specified in

Section 7.2.2 and any additional facility-specific design bases not specified

in the general system requirements.

REV. O. 2V6 7-13 - - STANDARD FORMAT AND CoNENT
CHAPTE7

* System description of each ESF actuation system similar to that specified in

Section 7.2.3. The description should include:

- any additional facility-specific system design

- features of the individual initiation and actation systems which

provide for them to function in concert to prevent or mitigate the

consequences of postulated accidents

* System performance analysis:

- an analysis of the operation and performance of each ESF actuation

system similar to that specified in the general system requirements

of Section 7.2.4, including analysis of the designs of any facilityspecific features or aspects

- a discussion of an analysis of the operation and performance of the

individual systems which allow them to function in concert to

prevent or mitigate the consequences of postulated accidents

- the bases of any technical specifications, including surveillance tests

and intervals specific to the design and operation of the subsystem

* Conclusions about capability, operability, and suitability of the ESF

actuation systems requested in Section 7.2.5.

7.6 Control Console and Display Instruments

Control console and display instrument systems and equipment include displays for

the reactor operator to view such operating information as current values of

operating parameters and the status of systems and equipment. The system also

enables the operator to control the reactor.

The applicant should describe how the control console and display instruments

have been designed to collect and display the operating information in such a

manner that it can be readily observed and interpreted by the operator. It should

describe how the manual control inputs (pushbuttons, switches, and other

equipment) have been grouped, oriented, and located with respect to the relevant

display instruments to enable the operator to best observe and interpret the

operating information and thereby take prompt and accurate steps to supply

control inputs on which the reactor control systems can act. In addition, the

combined and integrated functioning of the control console and display system

should be described to demonstrate how major equipment is designed to function

as an integrated information-handling system to readily aid the operator in

NUREG.1537, PART I 7-14 REV 0,2/96 NUREG-1537,PARTI 7.14 REV 0, 2/96
INmuMENrATiON AND CoNTROL SYSTMS

controlling operation of the reactor. The control console design should prevent

unauthorized operation of the reactor.

The advancement of digital technology has simplified the ability to gather, analyze,

manipulate, and display large amounts of data. A number of licensees have.

considered adding internally developed operator information display systems and

operating aids to their I&C systems.:If these systems digitally process control

console information and present this information to the reactor operator to inform

the operator of the status of the reactor, or if the operator uses such information to

make decisions about the operation of the reactor, the systems need to go through

the same review, including verification and validation of software as a digital RCS

or RPS. It is acceptable to locate these systems in areas where they cannot be

viewed by the reactor operator. The applicant should ensure that any interface

between the information display system and the control console is isolated.

The SAR system design criteria and basis information should include a system

description and a system performance analysis for each instrument system or major

equipment connected to or displayed at the control console. -The description and

analysis should be similar to those specified in Section 7.2 and should address the

following: -

* the outputs, controls, and operator interfaces

* how the output instruments are placed and how they are related to the

reactor and other system controls in the main console and auiiy control

room racks

drawings or photographs showing the arrangement of the display

instruments and console control equipment., -

* sufficient reactor-specific information for operators to understand functions

of both analog and digital systems, including connections and interaction

between them, and both redundancy and diversity of such systems

* the conclusion about operability and suitability for human factors as

requested in the general system recommendations of Section 7.2

7.7 Radiation Monitoring Systems

Radiation monitoring instrument systems should be designed to perform several

important diverse functions in the operation of a non-power reactor. These

monitors should indicate radiation intensity and may be used for reactor operations

such as to indicate the following: low coolant level, the need to actuate

containment or confinement systems, and the need for personnel radiation

REV. 0,2/96 7-15 ST4 uw FORMAT CONTENT

REV. O. 2/96' 7-15 STANDARD FORMAT AND CoLT~
CHAPTER7

protective actions, and to monitor release of radioactive material to the

environment. These systems include area radiation monitors, with displays near

the instrument location and in the control room. These systems may monitor

radioactive effluents in the form of gases, liquids, and airborne particulates and

provide continuous air monitoring (CAM) for airborne radioactivity in occupied

spaces such as the reactor room. Portable radiation monitors and personal

dosimetry systems should also be included to help assess exposure and prevent

overexposure of workers and other personnel. The radiation protective

instruments and measures should be discussed in detail in Chapter 11, 'Radiation

Protection Program and Waste Management." The present chapter should

concentrate on the I&C aspects of the radiation monitoring systems and should be

coordinated with the information in Chapter 11.

The applicant should briefly sunmarize the radiation-monitoring I&C system for

the facility and list the various systems and types of equipment. Since some of the

systems may provide input to the RPS or ESF actuation system, radiation

monitoring systems should meet the applicable criteria and requirements in Section

7.2 for those systems.

For each radiation monitoring system planned for the facility, the applicant should

give the I&C system design-basis information, a system description, a system

performance analysis, and a conclusion about the suitability of the system to

perform its function as specified in Section 7.2.

7.8 Bibliography

American National Standards Institute/American Nuclear Society, ANSI/ANS

10.4, "Guidelines for the Veification and Validation of Scientific and Engineering

Computer Programs for the Nuclear Industry," ANS, LaGrange Park, Illinois,

1987.

American National Standards Institute/American Nuclear Society, ANSIIANS

15.15, "Criteria for the Reactor Safety Systems of Research Reactors," ANS,

LaGrange Park, Illinois, 1978.

American National Standards Institute/American Nuclear Society, ANSVIANS

15.20, "Criteria for the Control and Safety Systems for Research Reactors" (draft),

ANS, LaGrange Park, Illinois.

American Nuclear Society, 'Transactions of the American Nuclear Society,

Session on Digital Control of Nuclear Reactors," pp. 248-259, Vol. 64, San

Francisco, California, November 10-14, 1991.

NUREG-1537,PARTI 7-16 REV 0, 2/96
INSTRUTATION AND CONTROL SYSTEM

Institute of Electrical and Electronics Engineers, IEEE Standard 74.3.2, "IEEE

Standard Criteria for Digital Computers Systems in Safety Systems of Nuclear

Power Generating Stations," Piscataway, New Jersey, 1993.

KEy. U, ;uytj 7-17 STANDMW FORMAT AND COInTNT

KEV. U. ZM9 7-17 STANDARD FoRmAT AND CON-=
Appendix 7.1

Regulatory Guide 1.152, Revision 1

Criteria for Digital Computers in

Safety Systems of Nuclear Power Plants

January 1996

STANDARD FORMAT AND CONTrNT
U.S. NUCLEAR REGULATORY COMMISSION Revision 1 -.Iamary 1996

REGULATORY' GUIDE

OFFICE OF NUCLEAR REGULATORY RESEARCH -

REGULATORY GUIDE 1.152

.(a was Issued as DG-1039)

- CRITERIA FOR DIGITAL COMPUTERS IN

SAFETY SYSTEMS OF NUCLEAR POWER PLANTS

A. INTRODUCTION

Criterion 21, Protection System Reliability and

Testability.' of Appendix A. 'General Design Crteria

for Nuclear Power Plants,' in 10 CFR Part 50, 'Domestic Licensing of Production and Utilization Facilities,' requires, among other things, that protection systems be designed for high functional reliability commensurate with the safety functions to be performed.

Criterion III, 'Design Control,' of Appendix B.

'Quality Assurance Criteria for Nuclear Power Plants

and Fuel Reprocessing Plants,' of 10 CFR Part SO requires, among other things, that quality standards be

specified and that design control measures be provided

for verifying or checking the adequacy of design.

''This regulatory guide describes a method acceptable to the NRC staff for complying with the Commission's regulations for promoting high' functional rehability and design quality for the use of digital computers in safety systems of nuclear power plants. The term

computer' is a system that includes computer hardware, software, firmware, and interfaces.

The Advisory Committee on Reactor Safeguards

has been consuled concerning this guide and has concurred in the regulatory position.

Any information collecton activities mentioned in

' this regulatory guide are contained as requirements in

10 CFR Part SO, which provides the regulatory basis

for this guide.' The information collection requirements

in 10 CFR Part 50 have been approved by the Office of

Managernient and Budget. Approval No. 3150-0011.

B. DISCUSSION,

Instrumentation and Control (I&C) 'systems that

use digital computers in safety systems make extensive

use of advanced technology, i.e.. equipment and design practices that are expected to be significantly and

functionally different from current designs. These designs include, but are not limited to, the use of microprocessors, digital systems and displays, fiber optics,

multiplexing, and different isolation techniques to

achieve the needed independence and redundancy.

IEEE Std 7-4.3.2-1993, "Standird Criteria for

*Digital Computers in Safety Systems of Nuclear Power

Generating Stations." I was jointly prepared by the Nuclear Power'Engineertng Committee of the Institute of

Electrical and Electromcs Engineers (IEEE) 'and the

Nuclear Power' Plant Standards Committee 'of the

American Nuclear Society (ANS). The NRC staff has

worked with IEEE and ANS, in- developing -IEEE

Std 7-4.3.2-1993 to ensure that the'guidance provided by the consensus standard is consistent with the

Commission's regulations. IEEE Std 7-4.3.2-1993

1IEEE publicadons may be purchased from the IEEE Seace

Center, 445 Hoes Lane. Piscataway, NiJ 0854.

11SMC REGULATORY GERlDES

Regulatory Guldes we lassue t OWNaUMb wad mailm available to thu pi~he wAihkftrmatlcn aso nutl oamPtable tolfwNs bOlitatIforhO h~pwmun- big ePaC111o Pars; of flu Cmmyvilsocs reou'tIOUs. taeflw~nsum tod by flustll vallabigSpcite n ~ieUor poctlaatellsowldfts.and,

llchas.i wasulatwy id as are mo 5.tsrtion o coreimarstlons atdmm-t

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enoustsad at all lI aus. and guldas; will be lavised. as afloopriale. to accommod0ate cornmrunts a to rerbctto Information or eywierlesa.

Wealtte ccmmiamts may be sutilttd toht Rules Review SWd DilectiS6s - B0w~ DFPS. AI, U S Nclear Regulaty Ccnvnslon. Washing- tn IDC 20ss.-ooo. .ll

11u guides ar Isud I theflos, n blaee tilaonta

1. P~ coaoru'. 6.P t

2 Asarc anid Taet Rectors - - 7 Trensportation 3. Fuels ar shtaetlis Facilities -8 OXMotI Health

,4 Eralivnmuntal an SlBlIft -" 9 Pot Pimicial t'ust.wid Anolw#

5 Materials and Plant Protection 10 Garl

S eope e t maybetltreeochargebywroit- kg the Office of Adrlnistratiol. Attention Distributen and Srovies Sectbn. U S. Nulear Regulatory Coinlaslon. Washing ton. DC 2(05550t:; or by fa at (3Ol4 260. - -

issued ouides "my also be poacased tron teu National Technical ilfte- matlon Serve onstandu order basi Detalls ontsrvice may be

ctaknbd by writlit NTIS.5265 Pont RoyalRoad, MISrngfIed. VA 22161

A- ; , ... 
has evolved from ANSI/IEEE-ANS-7-4 3.2-1982,

'Applications Criteria for Programmable Digital Computer Systems in Safety Systems of Nuclear Power

Generating Stations." IEEE Std 7-4.3.2-1993 is a significant improvement over its 1982 version. The 1993

version was approved by the IEEE Standards Board on

September 15. 1993. This standard identifies guidelines for digital computers (including hardware, software, firmware, and interfaces) to supplement IEEE

Std 603-1991. 'Standard Criteria for Safety Systems

for Nuclear Power Generating Stations "l The NRC

staff recognizes that development processes for computer systems continue to evolve.

Digital I&C systems share data transmissions.

functions, and process equipment to a greater degree

than analog systems Although this sharing forms the

bases for many of the advantages of digital systems, it

also raises a key concern with respect to its vulnerability to a different type of failure. The concern is that a

design using shared data bases and process equipment

has the potential to propagate a common cause failure

of redundant equipment. Another concern is that software programming errors can defeat the redundancy

achieved by the hardware architectural structure. Because of these concerns, the NRC staff has placed significant emphasis on defense-rn-depth against propagation of common cause failures within and between

functions.

The principle of defense-in-depth is to provide

several levels or echelons of defense to challenges to

plant safety, such that failures in equipment and human errors will not result in an undue threat to public

safety. A detailed defense-i-depth study and failure

mode and effect analysis or an analysis of abnormal

conditions or events should be made to address common cause failures. The Commission's position for

providing defense against common cause failures in

digital I&C systems for future light-water reactors is given in the Staff Requirements Memorandum of

July 21. 1993. on SECY-93-087, 'Policy, Technical,

and Licenising Issues Pertaining to Evolutionary and

Advanced Light-Water Reactor (ALWR) Designs"2

(specifically in point 18: 11 Q. 'Defense Against

Common-Mode Failures in Digital Instrumentation

and Control Systems').

Section 5.15, 'Reliability," of EEE Std

7-4.3.2-1993 states. 'When qualitative or quantitative reliability goals are required. the proof of meeting

the goals shald include software used with the hardware." The staff does not endorse the concept of

quantitative reliability goals as a sole means of meeutng

2Coptes at aYailable for Inspection or copying for a fee from the

NRC Public Document Room at 2120 L Street NW.. Washing- ton. DC. the PDR's mailing address is Mail Stop LL-6. Wasbington. DC 20555. telephone (202) 634-3273; fax (202) 634-3343

the Comm rson's regulations for reliability of the digital computers used in safety systems. The NRC staff's

acceptance of the reliability of the computer system is

based on deterministic criteria for both the hardware

and software rather than on quantitative reliability

goals.

Software failures that are not the consequence of

hardware failures are caused by design errors and,

therefore, do not follow the random failure behavior

used for hardware reliability. The NRC staff believes

that quantitative reliability determination, using a combination of analysis, testing, and operating experience,

provides information regarding the safety importance

of the computer system and also provides an added

level of confidence in its reliable performance. If

quantitative software relhability goals are used, the staff

believes that the amount of testing of the safety system

instrumentation and control equipment will increase.

The staff recognizes that the commercial dedication of

'commercially" available digital systems in nuclear

applications relies a great deal on quantitative methods

because of the operating experience data (such as

number of hours of successful operation) accumulated

over the years. The staff does not intend to preclude

operating experience data from the justification of a

successful commercial dedicauon.

Section 6, 'Sense and Command FeaturesFunctional and Design Requirements,' of IEEE Std

7-4.3.2-1993 indicates that no requirements beyond

IEEE Std 603-1991. are necessary. IEEE Std

603-1991 specifies the need to ensure acceptable response time for the instrumentation and control system

in order to accomplish necessary safety functions.

Consideration of the sampling rate of plant variables is

an important aspect of the design of a digital system

when satisfying this criterion.

IEEE Std 7-4.3.2-1993 icludes 8 annexes. This

standard states that these informative annexes are not

pan of IEEE Std 7-4.3.2-1993. The NRC staff believes these annexes contain information that may be

useful.; However, the information in these annexes

should not be viewed as the only possible solution or

method. Since a consensus has not been reached in

the nuclear industry, these annexes are noi endorsed

by the NRC staff.

C. REGULATORY POSITION

Conformance with the requirements of IEEE Std

7-4.3.2-1993, -Standard Criteria for Digital Computers in Safety Systems of Nuclear Power Generating Stations." with the exception of relying solely on quantitative reliability goals (Section S.15), is a method acceptable to the NRC staff for satisfying the Commission's

regulations with respect to high functional reliability

and design quality requirements for computers used as

components of a safety system.

1.152-2
Section 2 of IEEE Std 7-4.3.2-1993 references

several industry codes aind itandards. If a referenced

standard has been separately incorporated into the

Commission's regulations, licensees and applicants

must comply with the standard as set forth in the regulation. If the referenced standard has been endorsed

by the NRC staff in a regulatory guide, the standard

constitutes an acceptable method of meeting a regulatory requirement as described in the regulatory guide

If a referenced standard has been neither incorporated

into the Commission's regulations nor endorsed in a

regulatory guide, licensees and applicants may consider and use the information m the referenced standard

if appropriately justified, consistent with current regulatory practice.

D. IMPLEMENTATION

The purpose of this section is to provide information to applicants and licensees regarding the NRC

stafs plans for using this guide.

Except in those cases in which an applicant or licensee proposes an acceptable alternative method for

complying with specified portions of the Commission's

regulations, the methods described in this guide will be

used in the evaluation of submittals in connection with

applications for construction permits and operating licenses. It will also be used to evaluate submittals from

operating reactor licensees that propose system modifications voluntarily initiated by the licensee if there is a

clear nexus between the proposed modificauons and

this guidance.

VALUE/IMPACT STATEMENT

A draft Value/Impact Statement was published with the draft of this guide.

Task DG-1039. when it was published for public comment in May 1995. No

substantive changes were necessary, but a few editorial changes were made for

clarity and consistency. A copy of the revised ValuelImpact Statement for Revision 1 of Regulatory Guide 1.152 is available for inspection or copying for a fee

from the NRC Public Document Room at 2120 L Street NW.. Washington. DC;

the PDR's mailmg address is Mail Stop LL-6, Washington, DC 20SS5 telephone

(202)634-3273; fax (202)634-3343.

Federal Recycling Program

1.152-3
Appendix 7.2

Amendment No. 29

to Facility Operating License

No. R-38

General Atomics

STAemmRD FoRMAT AD CoNiNr
* °UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON. D. C 20555

> *'$ October 4, 1990

Docket No. 50-89

Dr. Keith A. Asmussen, Manager

Licensing, Safety and Nuclear Compliance

General Atomics

10955 John Jay Hopkins Drive

San Diego, California 92121-1194

Dear Dr. Asmussen:

SUBJECT: ISSUANCE OF AMENDMENT NO. 29 TO FACILITY OPERATING LICENSE

NO. R-38 - General Atomics

The Commission has issued the enclosed Amendment No. 29 to Facility Operating

License No. R-38 for the General Atomics TRIGA Mark I research reactor. The

amendment consists of changes to the Technical Specifications in response to

your submittal dated July 19, 1990.

The amendment approves the installation of a microprocessor based lnstrumentdtion and control system. The Technical Specifications are amended to reflect

the new system.

A copy of the related Safety Evaluation supporting Amendment No. 29 is enclosed.

Sincerely,

- *

Marvin M. Mendonca, Senior Project Manage

Ron-Power Reactor, Decommissioning and

Environmental Project Directorate

Division of Reactor Projects - III,

IV, V and Special Projects

Office of Nuclear Reactor Regulation

Enclosures:

1. Amendment No. 29

2. Safety Evaluation

cc w/enclosures:

See next page
4 UNITED STATES

$ ,NUCLEAR f + REGULATORY COMMISSION

WASHINGTON. 0 C 2065

GENERAL ATOMICS

DOCKET NO. 50-89

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 29

License No. R-38

1. The Nucledr Regulatory Cozmission (the Coamission) has found that:

A. The application for amendment to Facility Operating License No. R-38

.'filed by General Atomics (the licensee), dated July 19, 1990, complies

with the standards and requirements of the Atomic Energy Act of 19S4,

as amended (the Act), and the Commission's regulations as set forth in

10 CFR Chapter, I,

D. The facility will operate in conformity with the application, the

provisions of the Act, and the regulations of the Commission;

C. There is reasonable assurance:. (1) that the activities.authorized

by this amendment can be conducted without'endangering the-health

and safety of the public, and (1i) that such activities will be

conducted in compliance with the Cowtission's regulations set forth

in 10 CFk Chapter 1; r a s forth

D. The issuance of this amendment will not be inimical to the conmon

defense and security or to the health and safety of the public;

E. The Issuance of this amendment is in accordance with 10 CFR Part 51

of the Commission's regulations and all applicable requirements have

been satisfied; and

F. Prior notice of this amendment was not required by 10 CFR 2.105(a)(4)

and publication of notice for this amendment is not required by

10 CFR 2.106(a)(2).
-2-

.. Accordingly, the license is amended by changes to the Technical

Specifications as indicated in the enclosure to this license amendment,

and paragraph 2.C.(2) of License No. R-38 is hereby amended to read as

follows:

(2) Technical Srecifications

The Technical Specifications contained in Appendix A, as revised

through Amendment No. 29, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the

Technical Specifications.

3. This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Seymour M. Weiss, Director

Notn-Power Reactor. Decoaocissioning and

Environmental Project Directorate

Division of Reactor Projects - III,

IV, V and Special Projects

Office of Nuclear Reactor Regulation

Enclosure:

Appendix A Technical

Specitications Changes

Date of Issuance: October 4, 1990
ENCLOSURE TO LICENSE AMENDMENT NO. 29

FACILITY OPERATING LICENSE NO. R-38

DOCKET NO. 50-89

Replace the following page of the Appendix A Technical Specifications with

the attached pages. The revised pdges are identified by amendment number and

contain vertical lines indicating the areas of change.

Remove Insert

16 16

17
8. A sunmary of radiation exposures received by

facility personnel and visitors, including the dates

and time of significant exposure, and a brief summary of the results of radiation and contamination

surveys performed within the facility.

16 Amendr)ent No. 29
-T=I

)MXI REN= SA=~ SYSM SRAMS

originating Pbde in 1fich

Channe Set Point Effective

SS Pulse

1. Poer Level 275 kw or lower X

(2 indep nt channels) (1)

2. Fuel Elent 7tperature As specified in Section 7.4 X

(2 independent channels) or lower

3. Console Scram )anual X X

4. Facility Pwer Supply Supply Failure X x

5. Iagnet Current Key Switch Manual X X

6. Watchdog Tmer(2) Software Failure X X

Of) the rinimin required two inde-pendent channels, no more than one channel

shall utilize digital processing of pocer detector signals.

(2) 5hese scrams are only applicable when ccnputers are utilized to perfozm

reac-tor control functions.

E n

NII s slRU

I

Action Prevented

kImde in Which

Effective

SS Pulse

I. Withdrawal of zore than one standard rod x

2. Withdrawal of any standar-d rod X

3. Application of air to transient rod unless its rOvable X

cylinder is fully down

17 Amendment No. 29
UNITED STATES

<t NUCLEAR REGULATORY COMMI=SION

- WASHINGTON. D C. 20655

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

SUPPORTING AMENDMENT NO. 29 TO

FACILITY OPERATING LICENSE NO.'R-38

GENERAL ATOMICS

DOCKET NO. 50-89

1.0 INTRODUCTION

General Atomics (GA) has determined.that due to the obsolescen'ce and progressive

deterioration of their control console, .a'new reactor instrumentation and control

system is needed to maintain reliable operations. 'In December 1988, GA'published

their safety analysis of the new reactor instrumentation and control system. In

.this report GA concluded that the new system was anaillowable change'under

10 CFR 50.59.,- 10CFR 50.59 permits licensees to make changes in- tho facility as

described in the safety analysis report without prior Coamission approval unless

the 'proposed change, test, or experiment involves a change in the"Technical

Specifications incorporated in the license or an unreviewed safety question.

A proposed change, test, or experiment. shall be deemed to involve an unreviewed

safety~question (i) -If the probability-of occurrence or' the consequences'of an

accident'or.malfunction of equipment-Important to safety previously'evaluated in

the safety analysis report may be increased; or (ii) if-a possibility'for-an

accident or malfunction of a different type than any evaluated previously in the

safety analysis report may be.created; or (lii) if the margin of safety as defined

in the basis for.any technical specification is reduced.'

The staff concluded from.its review of the GA safety analysis-report that -

NRC review-and approval of the replacement computerized control system was"

required, since. () the installation of the new reactor instrumentation' and

control-system did present an unreviewed safety question because of the'possibility

of an accident or malfunction of a different.type than any evaluated previously

and (2) changes to the Technical Specifications were required.

Pursuant ;to 1O CFR 50.90, the licensee submitted-by.letter dated July 19,

1990, a'request to amend Appendix A of Facility Operating License No. R-38,

-Technical Specifications for theTorrey Pines TRIGA Reactor.0 -The licensee's

submittal.of July 19, 1990' included -the December 1988 safety analysis. The . requested'amendment would (1) allow'.installation of the micro-processor based

Instrument and'control systemn.(2) add the watchdog (software failure) scram

to Table' 1of the-Technical Specifications, Ninimum Reactor Safety System

Scrams"', and (3)'idd a requirement that no more than one of the required

two independent power level scram channels in Table 1 be a digital scram

channel.

The licensee has installed, in parallel to their existing control console, the

new digital microprocessor based instrumentation and control system. The

transfer of control from the old to the new system (including scram) was via a
-2-

-'I

series of gradual steps accompanied by tests which demonstrated the reliability

of the new equipment while maintaining the proven performance of the existing

control system. Upon completion of all testing (described later in this SER3,

the new console was used to control (except for the hardwired trip functions)

both the safety and nonsafety aspects of operation of the TRIGA reactor and

the old analog console was disconnected. The new console replaced the old

analog console in the control room. The primary functions of the new system

remained the same as the old system: to monitor critical parameters and

provide a scram signal when needed, to provide information to the operator and

to provide control for the pulse and steady-state modes of operation.

2.0 HARDWARE AND SYSTEMS ASSESSMENT

This portion of the review focused on the areas of potential vulnerability or

susceptibility of the new control console which might compromise its ability

to present accurate information to the operator and to provide scram signals

when required. ho assessment was made of the reliability of the nonsafetyrelated controls. Issues investigated included single failure, environmental

cLalification, seismic qualification, surge withstand capability (SWC).

electromagnetic interference (EVI), failure modes and effects, reliability,

error detection, and independence.

The primary review criteria for instrument dnd control systems for research

reactors are presented in ANSI/ANS 15.15-(1978) GCriteria for the Reactor Safety Systems of Research Reactors." The staff performed this evaluation also using criteria which apply to currert vintage nuclear power plants. However, due to

the inherent reactivity insertion safety feature of the TRIGA reactor design and

minimal decay heat generation that reduce the probability of fuel damage to a

minimum; the staff has concluded that these power plant criteria may serve as

guidelines and that strict adherence to the power plant criteria Is generally not

warranted. The exceptions are noted in the appropriate sections below.

During the review, the licensee described the new system including'licensing, engineering, testing and training aspects. The staff also had benefit of

material from the U.S. Air Force, the University of Texas at Austin and the

console owners group, as well as an independent safety review performed by

ORI, Inc. which concluded that the system was acceptable. The system for

GA's Mark I reactor is a similar system to that reviewed and approved in the

'Issuince of Amendment' No. 19 to Facility Operating License No. R-84 - Armed

Forces Radiobiology Research Institute" (AFRRI).

Similar to AFRRI, the GA Safety System Scram Circuit consists of two analog nuclear power monitor channels (NP-1000, and NPP-1000) and two fuel 'temperature channels which are hardwired. Different'from AFRRI,' the' NH-1000 microprocessor based nuclear power channel that monitors reactor power is wired to the scram

circuit and provides input to rod block. Also, wired into the scram circuit

at GA are contacts for manual scrdm, facility power supply failure scrasa, key switch scram, and watchdog (software failure) scram. Further, although not

required by Technical Specifications, there are scram features on (1) detector
- 3 -

high voltage failure on any one, power channel, (2) -loss of ac power to the

Instrumentation and ControlSystem due to earthquake switch trip, (3) externally

generated conditions, and (4) reactor power. reaching 1100 KW during a pulse.

2.1 Environmental and Seismic Qualification

The new control system is-installed in the control.room and the reactor hail.

The staff.considers the reactor hall to be avmild environment when compared to

power plant requirements and therefore the entire system can be considered to

be 1n'a mild environment.'. The system has been constructed in standard

coamerclal enclosures suitable for'a mild environment. The testing ar.d

operations, to date, have not revealed arty problems related to temperature

or humidity. The new system should not be unduly~susceptible' to teimperatureor humidity problems and is therefore acceptable tc the staff.

Though there have'.been no requirements promulgated for seismic'qualification

testing-of research reactor'control equipment, the staff considered the

equipment to determine gereral ruggedness. The licensee indicated that the

equipuent is mounted in a comr.ercial quality fashioi.whch -should prevent any significant movement of components vithin the console -and racks. : In this

TRIGA reactor,'an inrovertent scram does not present d sionificart challenge to reactor safety systems because a scrdm consists ofthe r&.oval of current

to the control rod magnets allowing the control rods to drop into the core by

,gravity;,andrlio other equipment is required to maintain the reactor Uiza safe

shutdown condition. The primary concern remaining would be relay.contact

chatter which'could preventai scram when required..-The safety system scram

circuits 'for'this'system are designed to scram on failure (which includes

contact chatter)'and therefore the staff concludes that any further testing is

not warranted and the system is acceptable...

2.2 Electromagnetic Interference (EMI)

The staff reviewed the susceptibility of-the new equipment to EMI due to the

potential for common mode interference which could disable.more than one

system at'a time. As discussed earlier, due to the design'characteristics.of

the.TRIGA reactor, an inadvertent scram does not-present a-significant challenge

to safety systems, though it might cause operational difficulties such as

disrupting an 'experi ment.

Industrial-type isolators'are generally used which prevent conducted EMI from

being transmitted between the control.and'safety mechanisms.. The neutron flux

signal cabling is shielded to'ireduce the impact of radiated EMI. Previous

experience with 'similar equipment"provided by several.different-vendorsvat

other. facilities"has indicated-that if -EMI causes any perturbance in the'-

system'it will most likely caus'e'a scram, which as previously discussed.1s-not

a safety concern. Based'on'the above,' the staff concludes that-.EMI should not

prevent a scram whEn required and the design is therefore acceptable.. .

2.3 Power Supplies

The power supplies for the'system are buffered to reduce the possible impact. of mrinor power line fluctuations.'. The scram circuits for the new system are
- 4

designed to scram when power is lost to them.' The NP-1000 and NPP-1000 are

analog devices and will respond to power fluctuations similar to the existing

analog equipment. The digital NN-1000 nuclear power channel uses a battery

backed-up random access memory (RAM) to store constant data during loss of

power. In addition to self-diagnostics, the 1NY-1000 has a watchdog timer

circuit which puts the KM-1000 in a tripped condition and scrams the reactor if

power fluctuations prevent proper software operation. As described in the

KM-1000 Software Functional Specification and Software Verification Program

(March 1989), the tH-1000 is also tested to verify that the system returns to

proper operation following restoration of power. The staff finds this accept- able.

2.4 Failure Modes and Effects

The December 1988 safety analysis included Scram Circuit Safety Analysis performed

by the University of Texas at Austin. This'study identified the various ways in

which the reactor safety system could fail. These include:

1) Physical System Failure (wire breaks, shorts, ground fault circuits)

2) Limiting Safety System Setting Failure (failure to detect)

3) System Operable Failure (loss of monitoring)

4) Computer/Manual Control Failure (automatic and manual scram).

This study was based on a fault tree approach which predicted failure to scram

for various failure modes. The study concluded that a failure of all safety

systems and therefore failure to scram was extremely unlikely. Failures

attributable to the unique failure modes of the software of the NM-1000 were

considered. The staff concludes that the failure modes and effects of

the new system were acceptably addressed.

2.5 Independence, Redundancy and Diversity

The staff reviewed-the data link between the safety channels and the nonsafety

systems. The safety'channels provide direct hard-wired scram inputs and are

also hardwired directly-to independent indicators on the control console.

The operators are-provided with information from both the analog NP-1000 and

NPP-1000 power monitors and the digital NM-1000 monitor. The information is

displayed on both direct wired bar graphs and on a graphic CRT. In addition,

the safety channels provide inputs to the Non-Class lE Data Acquisition

Computer (DAC) through isolators. The isolators used have not been tested

for maximum credible faults which the staff'requires for power plant use,

but have been tested by the manufacturer to standard commercial criteria.

The DAC is then'connected'via'redundanthigh speed serial data trunks to the

Non-Class 1E Control System Computer (CSC) which interfaces with the operator

by controls, a keyboard and CRT displays. 'Since the CSC does communicate with

the safety channels, this aspect of the system would not meet the independence

requirements of a power plant, but the staff concluded it was not necessary

for the current application at GA.

Further, the scram circuit is essentially unchanged in that it maintains the

fail safe design using the same automatic and manual contacts which open to

remove power to the control rod magnets. For the GA facility, redundant fuel

temperature inputs are provided to the scram circuit. Redundant power level

inputs (NP-1000, NPP-1000) to the scram circuit are also provided.
- 5 -

This system has also added the computer watchdog scram and the digital

WM-1000 scram. At GA, in additionto the KM-1000 being wired to the scram

circuit, It' provides inputs to the rod withdrawal prevent'interlock system.

The use of both analog and digital neutron monitoring, and the watchdog'scram

function provides additional diversity and redundancy to the scram system.

The system as installed meets most of the requirements of IEEE-279-1971:-

^Criteria for Protection Systems forNuclear, Power Generating Stations and

IEEE 379-1977- Application of, the Sirgle-Failure Criteria to Nuclear Power

Generating Station Class'lE Systems. .

The staff has concluded that the.level of Independence, redundancy and diversity

which has been maintained is acceptable for the GATRIGA reactor.

2.6 Testing'and Operating History.-

Extensive testing of the new system has been done by both GA and AFFRI. A

significant number of design changes took place during the testing and

phase-in of the new system. The staff has reviewed the problems discovered

during testing of the system and has concluded that the resolutions appear

acceptable. The staff also agrees with the licensee that long-termn - ';

operability and-safety is.enhanced due.to installation of-equipment which

has spare parts readily available. An additional improvement is the self::<

diagnostics feature which allows continuous on-line testing and reduces-the

possibility of undetected failures.

3.0 SOFTWARE ASSESSMENT

3.1 Criteria

The staff requires an approved verification and validation (M&V) plan for

software which performs a safety function or provides information to the operator. At GA,-theNM-1000 provides inputs to the scram circuit and the rod

withdrawal prevent interlock system block function. -The H4-1000 software

development was reviewed by the staff -to determine.the acceptability of the

V&V plan. The staff.compared the General Atomics V&Y plan to Regulatory

Guide 1.152."Criteria for Programiable Digital'Computer Software in Safety-.

Related Systems at Nuclear Power'Plants. which endorses AHSL.IEEE 7-4.3.2 -

1982 Application.Criteria.for Progrannable DigitalComputer Systems in Safety

'Systems of Nuclear Power GeneratingStations.'- The staff has concluded that.c

this standard is appropriate for use in reviewing research reactor software.

3.2 Verification and Validation Plan

The staff auditeidthe verification and validation'documentation provided by.;

General Atomics.': For~the installationat the GA.TRIGA the 10-1000 is wired.

directly intothe scram ,circuit, and therefore' requires highly reliable

software to perform-'its''safety function when required. The assessment of the
- 6 -

NM-1000 software built by General Atomics is an assessment of the methodology and procedures used to develop the software. The process is evaluatedby reviewing the verification and validation trail through the development

process.

Verification and validation (V&V) are two separate but related activities

that follow the development of software. Verification determines whether the

requirements of one phase of the development cycle have been consistently, correctly, and compIetely transformed (fulfill the requirements) to the

subsequent phase of the cycle. Validation is the testing of the final product to ensure that performance conforms to the requirements of the initial specification. The need for V&V arose because software is very complex, and prone to

human errors of omission, commission and interpretation. V&V provides for an

independent verifier to work in parallel with, but independent of, the development team to ensure that human errors do riot hinder the production of safety

software that is reliable and testable.

In executing Y&V, certain principles have proven over time to be very effective

in software programs:

- Well defined systems requirements expressed in a well written document. - Development methodology to guide the production of software. - Comprehensive testing procedures. - Independence of the V&V team frow the development organization.

These principles can serve as a comprehensive reference base for applying the

applicable criteria for software evaluations of Class lE safety systems, and

were used as guidance in the following review areas.

3.3 Independence

P. key ingredient in an effective verification process is the independence of

the verifier. For the NM-1000 the original' software was- developed by Sorrento

Electronics. After General Atomics obtained the rights to market the NM-1000

for research reactors, a software consultant was used to modify the'software.

After many changes had been made another contractor was brought in. Each

contractor in turn assured an additional level of independent review from the

original design. Though the-requirements imply a concurrent review the staff

finds that the verification has been sufficiently independent and is therefore

acceptable for research reactors.

3.4 Validation Testing

The validation testing must be done by a team that did not participate in the

design or implementation of the software product. General Atomics used the

Neutron Monitoring System Acceptance Test Procedure as parn of the validation

testing. In addition the staff reviewed substantial additional validation

testing which has been perforcied at the AFRRI fdcility. The staff did note a

functional description of unknown date which included samples of the computer

code. Though the people involved ir development knew the specific functions
- -7 -

which the 10-1000 was to perform these had never been written.down,.which allows

substantial possibilities for ooission when preparing test procedures. Upon

request from the staff, General Atomics provided the.functionial specification'- E117-l0Ol: (March 1989) which lists .in detail. the functionsperformed by the-' t

NM-1000. Included in this specification was a:cross reference where 'the

vendor-verified that each specific functional.requirement had been tested.-:

The staff finds that this testing and verification is acceptable.

3.5 Discrepancy Resolution. -

A key element in any verification and vdlidation effort is the process by which discrepancies uncovered.during development.are recorded, identified,

resolved and corrected. The resolution of a discrepancy must be.reflected'in'

all applicable documents (e.g., source code, the software design specification,

the software requirements,--and-the original systems specification). The staff

reviewed discrepancies and other-coavnents-provided to General Atomics by the

Console Owners Group and found that-the process anderesolution were documented

and appeared adequate. Whendiscrepancies resultedsin-code modification, a -'

description of the changes aria it's reason was added 'to the code annotation.;'

The staff finds the discrepancy resolution methods by General Atomics to'be -

acceptable. .. :. -

3.6: Design Approach. -

The primary, software specification provides the foundation for not only sound

oevelopment, but also for effective verification and validation activities.

The individual requirements in the specification for any software system

describe how the software is to behave, in any circuostance. .The specificationnust be reliable and testable. A reliable specification exhibits the following

characteristics:.

- Correct - Each requirement of the. safety, function has" been stated correctly.'. - Complete - All of the requirements for the safety, function are included. - - Consistent - The requirements are complementary and do not.contradict each.

other. , * . -

- Feasible -' The requirements can be satisfied with available technology. - Maintainability - The requirements will be satisfied for the lifetime of

the equipment. .. - Accurdcy - The requirements include the acceptable-bounds'of operation.'

The staff reviewed the design approach.with General Atomics. The documentation

was found to be -lacking-in several areas with.the most significant being the

lack of a functional requirements specification which GA has since prepared.- Documentation of-the early development was sketchy which was attributed

partially to the'transfer of-the product without including all of.the backup -:-

information..'.The documentation of recent changes has improved significantly. -

Though the staff finds that the design approach for the HM-1000 since inception

has been.erratic,.the recent, development work appears to' be improved in struc--

ture and control. - . . -
- 8 -

3.7 Software Evaluation

The software development plan for the 1H-1000 appears to the staff to be a

very specific design goal oriented'development, where the application and

basic hardware and software requirements were known by the designers; however,

there was no step by step plan developed. The failure to have a step by step

plan such as described in ANSI/IEEE 7-4.3.2 - 1982 resulted in the'need for

General Atomics to retrofit the functional requirements document and verify

that each requirement had been tested. To meet this requirement GA developed

the XM-1000 software verification program (EI17-1002 March 1989). The staff

also reviewed working copies of the NM-1000 design input which demonstrates

that the functional, requirements are currently well understood by the design

team and concludes that the software should perform its intended safety

function as required.

The staffs review indicated that GA could benefit through the development of a

corporate software 'development plan that can be applied to any future Class IE

software prior to starting design. The plan could include a description of

the development phases in sufficient detail so that the verification and

validation efforts can be initiated-at the beginning of any design effort.

The plan could also contain a taxonomy of documentation, and reviews which

demark the injection points for verification and validation activities. A

corporate software development plan for Class lE systems could prove to be

effective in the development of reliable software consistent with the intent of

ANSI/IEEE-ANS-7-4.3.2 - 1982.

3.8 Operator Task Analysis

In reviewing the documents it became apparent that there was not a formal

task analysis to support the design of the operator interface. The initial

specifications and descriptions were vague. After the equipment and software

were substantially designed, the functional requirements and working level

descriptions did include the operator task requirements. A task analysis

prior to development would probably have minimized the software iterative

process and therefore provided less opportunities for error. The staff

concluded that through the V&Y process the requirements have been specified,

and incorporated in the design. Therefore, the V&V plan is acceptable.

4.0 TECHNICAL SPECIFICATIONS

As previously discussed, the presentation of correct, timely information

to the reactor operator contributes to the safe operation of the reactor.

The scram circuit at GA will. include watchdog timer contacts which will

provide a scram upon software failure. Therefore to assure the presentation

of timely, correct information to operators or the proper safety system scram,

the watchdog scram inputs are added to Table I, Minimum Reactor Safety System

Scrams of the Technical Specifications. Additionally to assure acceptable

diversity of the new system, Table I has been amended to specify that of the

minimum required two independent power level channels, no more than'one

channel shall use digital processing of power detector signals.

I . .
- 9 -

5.0 ENVIRONMENTAL CONSIDERATION

This amendment involves changes in a requirement with respect to the installation

or use of facility components located within the restricted area as defined in

10 CFR Part 20. The staff has determined that the amendment involves no

significant increase in the amounts, and no significant change in the types, of

any effluents that may be released offsite, and there is no significant increase

in individual or cumulative occupational radiation exposure. Accordingly, this

amendment meets the eligibility criteria for categorical exclusion set forth in

10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact

statement or environmental assessment need be prepared in connection with the

issuance of this amendment.

6.0 CONCLUSION

The staff concludes that the hardware design of the new General Atomics

console is acceptable for use in the GA TRIGA reactor. The Software design in

the CSC, DAC and HM-1000 will not prevent the safety functions of the hardwired

scram circuit from performing and is therefore acceptable. The technical

specifications are amended to include the watchdog scram inputs and maximum

use of digital power measurement channels.

The staff has also concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the

probability or consequences of accidents previously evaluated, or create the

possibility of a new or different kind of accident from any accident previously

evaluated, and does not involve a significant reduction in a margin of safety,

the amendment does not involve a significant hazards consideration, (2) there

is reasonable assurance that the health and safety of the public will not be

endangered by the proposed activities, and (3) such activities will be conducted

in compliance with the Commission's regulations and the issuance of this

amendment will not be inimical to the common defense and security or the

health and safety of the public.

Principal Contributor: James C. Stewart

Dated: October 4, 1990
Appendix 7.3

Amendment No. 30

to Facility Operating License

No. R-2

Penn State Breazeale Reactor

ST&ND"M FORMAT AM COMNTIr
Do

NUCLEARa UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON. D C.

cket No. 50-5

August 6, 1991

Dr. Charles L. Nosler

Vice President for Research and

Dean of the Graduate School

Pennsylvania State University

207 Old Main Building

University Park, Pennsylvania 16802

Dear Dr. Hosler:

SUBJECT: ISSUANCE OF AMENDMENT NO. 30 TO FACILITY OPERATING LICENSE NO. R-2 -

PENN STATE BREAZEALE REACTOR

The Commission has issued the enclosed Amendment No. 30 to Facility Operating

License No. R-2 for the Penn State Breazeale Reactor. The amendment consists

of changes to the Technical Specifications in response to your submittal of

April 19, 1991 as supplemented on July 8, 1991.

The amendment approves the installation of a new reactor instrumentation and

control console system. The Technical Specifications are amended to reflect

this change.

A copy of the related Safety Evaluation supporting Amendment No. 30 is enclosed.

Sincerely,

Marvin M. Mendonca, Senior Project Manager Non-Power Reactors, Decommissioning and

Environmental Project Directorate

Division of Advanced Reactors

and Special Projects

Office of Nuclear Reactor Regulation

Enclosures:

1. Amendment No. 30

2. Safety Evaluation

cc w/enclosures:

See next page
C ofCLEAR UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON. C' 

PENNSYLVANIA STATE UNIVERSITY

DOCKET NO. 50-5

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 30

License No. R-2

I. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment-filed by the Pennsylvania State

University (the licensee),-dated April 19, 1991 as supplemented on

July 8, 1991, complies with:the standards and requirements of the

Atomic Energy Act.of 1954, as amended (the Act), and the Commission's

rules and regulations as set.forth in 10 CFR Chapter 1;

B. The facility will operate in conformity with the application,

the provisions of the Act, and the rules and regulations of the

Commission;

C. There is reasonable assurance: (i) that the activities authorized

by this amendment can be conducted without endangering the heilthand safety of the public, and (ii) that such activities will be

conducted in compliance with the Commission's regulations;

D. The issuance of this amendment will not be inimical to the common

defense and security or to the health and safety of the public;

E. The issuance of this amendment is in accordance with 10 CFR Part 51

of the Commission's regulations and all applicable requirements have

been satisfied; and

F. Prior notice of this amendment was not required by 10 CFR 2.105(a)(4)

and publication of notice for this amendment is not required by

10 CFR 2.106(a)(2).
- 2 -

2. Accordingly, the license is amended by changes to the Technical

Specifications as indicated in the enclosure to this license amendment,

and paragraph 2.C.(2) of License No. R-2 is hereby amended to read as

follows:

(2) Technical Specifications

The Technical Specifications contained in Appendix A, as revised

through Amendment No. 30, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the

Technical Specifications.

3. This license amendment is effective as of August 11, 1991.

FOR THE NUCLEAR REGULATORY COMMISSION

Seymour H. Weiss, Director

Non-Power Reactors, Decommissioning and

Environmentil Project Directorate

Division of Advanced Reactors and

Special Projects

Office of Nuclear Reactor Regulation

Enclosure:

Appendix A Technical

Specifications Changes

Date of Issuance:
ENCLOSURE TO LICENSE AMENDMENT NO. 30

FACILITY OPERATING LICENSE NO. R-2

DOCKET NO. 50-5

Replace the following pages of the Appendix A Technical Specifications with

the attached pages. The revised pages are identified by amendment number and

contain a vertical line indicating the area of change.

Remove Insert

7 7

14 14

15 15

16 16

17 17

31 31

32 32
7

reactive rod Is in Its most reactive postion. and ta the reactor wi remain

subcitacal without furher operator action.

1.1.42 SQUARF WAVE DERATION

Square wave (SW) operation sh mean operation of the reactar with the mode

selector switch In the square wave position which aiows the operator to Irsert

preselected reactivty by the ejection of the tansiern rod. and which resut hI a

maxdul power of 1 MWor kss.

1.1.43 . RIGAFUFt. LFMP-T

A TRIGA fuel elemert k a singe TRIGA Se rod of stardw type, ether 85 wt% U.

ZrH staInless steel cladding or 12 wt% U-ZrH In stailess steel cading enriched to

less than 20% uraniurni-235.

1.1.44

A watchdog cirwit Is a circztt consisting of a tiner and a relay. The titner energizes the relay as long as t is reset prto te ten of the brg intervaLt If is not

reset within the tinig hterval, the relay wlU de-energize thereby causing a SCRAM.

2.0 SAFETY LIMr AND LIMmNG SAFF SYSTE- SFTI1NC

2.1 SAFETY LUMrT-FVEL FLEMENTTEMPERATURE

Amlicabifity.. ... .

The safety mit specification applies to the mraxlrm terrperalure in the reactor fuel.

Ohipetive

The objecive is to define the makdrrum fuel element terrperature that can be

permitted with conridence that no damage to the fuel element and/or cladding will

result.

The temperature In a watermcooled TRiGA fuel element shall not exceed 1 150C

under any operatin condition.

The inportart parameter for a TRIGA reactor Is the fuel element terrperare. This

parameter Is well wited as a sinle specification especially since it can be measured at

a pont wfthin the fuel element The measured fuel terrperature is directly related to

the nmxiaxum fuel teqTnerature of the region. A loss In the integrity of the fuL element

cladding could arise from a bukiup of excesswe pressure between the fuel-moderator

and the cladding I the maxirmum fuel termpetatrure exceeds I 1509C. The pressure is

caused by the presence of air, fission product gases, and hydrogen from the

dissociation of the hydrogen and zirconium in the fuel-moderator. The magnitude of

this pressure is determnined by the fuet-moderatortemperature, the ratio of hydrogen

to zirconium in the alloy, and the rate change in the pressure.

Amendment No. 30
3.2 CONTROL AND SAFETY SYSTEM

3.2.1 REFACTOR CONTROL RODS

This specification apries to the reactor control rods.

The objective Is to assure that sufficient control rods are operable to mintain the

reactor subaitical.

S~ficatio

There shall be a mrinium of three operable control rods in the reactor core.

Basis

The shutdown margin and excess reactivity specifcaliorn require that the reactor can

be made subcitical with the most reactive control rod withdrawn. This specilcation helps assure Rt.

3.2.2 MANUAL AND AUTOMATIC CONTROL

ADI21cab;ilit

This specification applies to the maxnum reactivity insertion rate associated with

movement of a standard control rod out of the core.

The objective is to assure that adequate control of the reactor can be maintained

during manual and 1. 2. Or3 rod automatic control.

Sn~eedcf io}

The rate of reactivity insertion associated with movement of ether the regulating. shim.

or safety control rod shall be not greater than 0.63%0hkfIl (-90e) per second when

averaged over fun rod travel. II the automatic control uses a corbination of more than

one rod, the sum of the reactivity of those rods shall be not greater than 0.63%o Ak

(-90c) per second when averaged over full tnvel.

Bacsi

The ranip accident analysis (refer to Safety Evaluation, Chapter IX) hdicates that the

safety timit will not be exceeded if the reactivity addition rate is less than $2.50/second.

when averaged over fun travel. This specifcation of S0.90/second, when averaged

over fun travel, is well wttain that analysis.

Amendment No 30
Is

3R2.3 REACTOR CONTROLySTEM

tecation appfies to the Infomatbn which mist be avalable So the reactor

operator during reactor operaton.

The objective Is to require that8 sUffleru Information Is available to the operator to

assure saWe operation of the reactor.

Zoen D .

The reactor sh r not be operated unless e measuring channels isted in Tabe are

operabe. (Note that MN. AU. and SW are abbreviations for manual. automaic; and

square wave, respectively).

Table I

Mn. No Effective Mode

Measurino Chnl Operilblh MNU ei dz S

Fuel Element Temperature

Liear Power,

Percern Power '.

Pulse Peak Power

Count Rate

Log Power

Reactor Period

I

1

I

I

I1

I

I

x

x

x

x

x

x

x x , -. . 1 %.

x

x

x

Fuel temperature displayed at the control console gives continuous Information on this

parameter which has a specilied saety Itft. The power level moMors assure that the

reactor power level is adequately montored for the manual, automatc. square wave

and pulsing modes ot operation. The specifications on reactor power level and reactor

period indications are Included In this section to provide assurance that the reactor is

operated at all tmes within the limits asowed by these Technical Specifications.

AmendffwntNo 30
1 6

I

3.2=4 RFACTOR SAFTY SYSTEM AND INTERLOCKS

This spcfication applies to the reactor fety system channL, thot Wer"oc and the

watctwog cir

Tbe ojectie Is to specify Ute ninimurm number ot reactor saety sstem channels and

Interlocks that must be operable for stee operatin

The reactor sha not be operted unless a of the channed w and bterbcs descrbed in

Table 2a and Table 2b are operable.

Table 2a

I Minimum PSBR Channels

I Channel

Fuel Temperature

High Power

Number

1

2

Functio

SCRAM -C7OO'

SCRAMS II 0% ofIl

MW

Effective Mode

UNAU X

x x x

x x

i Detector Power Supply

Scram Baron Console

Preset Timer

I

I

I

SCRAM an taire of

supply voltage

Mma aum

Transient rod scram 15

secnds or less after

puls

x X

x x

x

x

I Watchdog CirctC 1 SCRAM on software or

s-dech faibure x x x

Amendment No. 30
17

Table 2b

Mkinirni PSB3R lrteocks I

Number

Ooerah e

Eftective Mode

Source Level I

Log Power

Transient Rod

Prever rod withdrawal

wi ess than two

neutron induced counts

per second on the

startup lannel

Prevent pulsing from

le"vs above I kW

Prevent applications of

air mless ylndrisfufly

Inserted

X

1

I

X

X

Shim. Safety, and

Regulating Rod

I Movemern of any rod

except transient rod

X

Sirnuttaneous Rod

WWtftwaI

1 Prevents sirnutaneous

mar&a wthdrawal of two

rods

X X

A temperature scram and two power level arms provide automatic protection to

assure that the reactor is sthA down before the safety Int on the tuel element

temperature wil be exceeded. The manual scram allows the operator to shut down the

system in any mode of operation I an unsafe or abnormns condition o05r. In the

event of faikure of the power supply for the safety chbors, operation of the reactor

Whout adequate is prevented. The preset timer Insures Wat the

transient rod wi be inserted and the reactor will remain at tow power after pung. The

watchdog cul wi scam the reacor If the software or the selfcheds fal (see Safety

Analysis ReponMChapter VIl sectlons H2d and L4)

In the pulse mode, movement of any rod except the transient rod Is prevented by an

Interlock. This Itherlock action prevents the addition of reactivity over that in the

transient rod. The intero to prevent startup of the reactor with less than 2 cps

assures tat sufficient neutrons re available for proper startup in an relevant modes of

operation. The hedock to prevent the inaton of a pulse above 1 kW Is to assure that

the magnitude of the pulse will not cause the safety lmit to be exceeded. The

interiock to prevent application of air to the transient rod unless the cylinder is fully

insened is to prevent pulsing the reactor in the manual mode. Simultaneous manual

withdrawal of two rods s prevented to assure the reactivity rte of insertion Is not

exceeded.

Amendme-o! No 30
31

.hsertion rates, and the reactivity worth of experiments Inserted in

. the core.

422 IREAC7VT NERON RAE

This spedfication applies to control rod movement speed.

The objectve Is to assure tha the readivity adftion rate speciication is rot violated

and ta the control rod drives re functioning.

Sneffratfan

- The rod drive speed both up and down and the time from sarm Initlation to the full

I insertion of any control rod from the full up position shall be measured annually. not to

exceed 15 months. or when any significant work is done on the rod drive or the rod.

;, . -Bask - .. -:;

This specification assures that the reactor will be prorptly shut down when a scran

signal Is initated. Experience and analysis have Indicated that for the range of

transients anicpated for a TRIGA reactor, the specdied scram time is adequate to

assure the safety of the reactor. It also assures that the maxmirru reactvi addItion

rate specification will not be exceeded.

42.3 REACTOR SAFTY AND CONTROL SYSTFMS

The specificalions apply to the surveillance requirements for measurements, channel |

tests, and ciannel checs of the reactor safety systems and watchdog circuit.

Obieclive

The objective is to verify the performance and operabiry of the systemns and

components that are directly related to reatorsafety.

--. a A charnel test of the sram function of the high power, fuel temperature, manual.

and present timer safety channels shall be made on each day that the reactor is to

be operated, orpriorto each operation that extends more than one day.

b. A channel test of the detector power supply SCRAM function and the watchdog

circut shall be performed annually, not to exceed 15 months.

Amendmom No 30
-

c. Channel checks for operabity shall be performed daily on fuel element

temperature linear power. count rate. log power and reactor period when the

reactor Is to be operated, oa prior to each operation that extends mome tan one

day.

d. The percant power chanel ha be compared wdh other hdependent channeLs

br proper chaniel bndicaion, when appropriate, each ime w reactor is

operated.

e. The pulse peak power channel shall be conpared lo the fwl teoperature each

tino the reactor is pulsed lo assure poper peak power channel operation.

TRIGA system components tihe poen operational renabty. Oaiy chanel tests

Insure ao&zre scram functions and ins the detection of possible charnel drift or

other possble deterioration of operating characetestis Tbe channel chedcs will make

Inormation availablo to the operator to assure sge operation on a daily basis or prior to

an extended nt. An anwal channel tes ol the detector power supply scram will

assure that this system works, based on past experience as recorded In the operaon log book. An anmal channel test of the watchdog ckvA Fs sufficient to assure

operabify. Comparison of the percent power channel wIM ofer ndependent power channels will assure the detection of charel drft or other posstble deterioration of its

operational characteristics. Comparison of the peak pulse power to the fuel

temperature for each pulse will assur the detection of possile channel drMt or

detenoration of its operational characteristics.

4Z4 ARACTOEI INTERLOCKS

ibis specification applies to the surveillance requWements for the reactor corrol

system iteroks.

The objecte is to inure performance and operably of the reactor control system

interlocks.

S cpfintios

L A charel check of the source interloc all be performed each day a the

reactor Is operated or pnor to each operaton that extends more than one day.

b. A channel test shall be performed semfiarnualy, not to exceed 7 months, on the

log power interlock wch prevents pulsing from power levels higher than one

kilowatt

Arnridmen No. 30

-
o *UNITED STATES

NUCLEAR REGULATORY COMMISSION . E WASHINGTON. D.C

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

SUPPORTING AMENDMENT NO. 30 TO

FACILITY'OPERATING LICENSE NO. R-2 .

PENNSYLVANIA STATE UNIVERSITY . ..

-OCKE ;;N.' 'H50-5 ; ;: -

1.0 INTRODUCTION .

The Pennsylvania State'University (the licensee) has determined that a new

reactor instrumentation and control console system should provide'improved

reliability, easier repair, and'increased information to the .operators.-

-Therefore, the licensee's-submittal of April 19, 1991 as supplemented on

July-8, 1991, requested 'changes to allow 'installation of the new-reactor,- --instrumentation and control console system. ' . ' . ;

2.0 DISCUSSION '

The transfer of 'control to the new 'system (including scram functio'ns) is to be

via a series of gradual steps' accompanied by tests'which demonstrate the , reliability of the new equipment while maintaining the proven performance of

the existing control system. Upon completion of. all testing (discussed later

in this'safety evaluation),'the new console will be used to control -(except

for the hardwlred trip functions) both the safety and non-safety aspects of

operation of the TRIGA reactor. The new console will-replace the old console

in the control room. The primary functions.of the new system remain the same

as the old system: to monitor criticdl parameters and provide a'scram' signal

when needed; to provide information'to the operator; and.to provide control

for operations.

2.1 Hardware and'Systems Assessment

This portion of the review focused on the areas of potential vulnerability or

susceptibility of the new control console.which might compromise its ability

to provide accurate information to'the operator and to provide scram signals

when required. -No assessment was made' of the reliability of the non-safetyrelated controls. Issues investigated included 'single failure, environmental

qualification,' seismic qualification, 'surge'withstand capability (SWC),

electromagnetic interference (EMI), failure modes and effects, and independence.

The primary review criteria for instrumentation and control systems for

research reactors are included in ANSI/ANS 15.15 (1978) "Criteria for the

Reactor Safety Systems of Research Reactors." The staff performed this

evaluation also using criteria which apply to current vintage nuclear power

plants. However, due to the inherent reactivity insertion safety feature of

the TRIGA reactor design and minimal decay heat generation that reduce the
- 2 -

probability of fuel damage to a minimum, the staff has concluded that these

power plant criteria may serve as guidelines and that strict adherence to the

power plant criteria is generally not warranted. The exceptions are noted

in the appropriate sections below.

2.1.1 Environmental dnd Seismic Qualification

The new control system will be installed in the control room and the reactor

room. The staff considers the reactor room to be a mild environment when

compdred to power plant requirements and therefore the entire system can be

considered to be in a mild environment. The system has been constructed in

standard commercial enclosures suitable for a mild environment. The testing

and operations, to date, have not revealed any problems related to

temperature or humidity. The new system should not be unduly susceptible to

temperature or humidity problems and is therefore acceptable to the staff.

Though there have been no requirements promulgated for seismic qualification

testing of research reactor control equipment, the staff considered the

equipment to determine general ruggedness. The licensee indicated that the

equipment is mounted in a commercial quality fashion which should prevent any

unacceptable movement of. components within the console and racks. The most

likely outcome'in any seismic event for the new control system would be a

reactor scram. In this TRIGA reactor, an inadvertent scram does not present a

significant challenge to reactor safety systems because a scram consists of

the removal of current to the control rod magnets allowing the control rods

to drop into the core by gravity. No other equipment is required to maintain

the reactor in a safe shutdown condition. The primary concern remaining

would be relay'contact chatter which could prevent a scram when required.

The safety system scram circuits for this system are designed to scram on

failure (which includes contact chatter) and therefore the staff concludes

that any further testing is not warranted and the. system is acceptable.

2.1.2 Electromagnetic Interference (EMI)

The staff reviewed the susceptibility of the new equipment to EMI for potential

common mode interference which could disable more than one system at a time.

Industrial-type isolators and input/output cards are generally used which

prevent conducted EMI from being transmitted between the control and safety

mechanisms. The console's signal cabling is shielded and where possible

separated from the power cabling to reduce the impact of EMI. Further, the

power supply is filtered for high frequency EMI. Based on the above, the

staff concludes that EMI should not prevent a scram when required and the

design is therefore acceptable. Previous experience with similar equipment

provided by several different vendors at other facilities has indicated that

if EMI causes any disturbance in the system it will most likely cause an

inadvertent scram, which as previously discussed is not a safety concern.
- 3 -

2.1.3 Power Supplies

The power supplies for the system are filtered and protected with surge withstand capabilities to reduce the possible impact of minor power line f~hctuations. The scram circuits for the new system continue to scram when power is

lost tb them. The control rod motor controllers use a battery back-up random

access memory (RAM) to store constant data during loss of power. In addition

to self-diagnostics, the computerized control system has watchdog timer ' ' circuits which scram the reactor if anything, including power. fluctuations,7

prevent proper software operation.' The control system is also tested to - verify that the-system returns to proper operation following restoration ofpower. The staff finds this acceptable. -

2.1.4 -Failure Modes and Effects

The licensee's safety analysis referenced *Failure Modes and Effects

Analysis," AECL Document FMA-17-60501-001,,:Rev. 0.-. This study identified the

various ways in which the reactor safety system could fail. The study concluded

that: "no single failure will prevent a reactor scram or:desirable interlock.

Although certain failures may impair a particular trip,-there islalways-an alternative trip parameter.... available to initiate a SCRAM." Further, the

licensee indicated in their safety analysis that in a postulated event of-the

complete failure of the reactor safety system coincident with the occurrence

of any analyzed accident, the radiological consequences would be negligible.

The staff concludes that the failure modes and effects of.'the new system were

acceptably addressed and even in the event of a unforeseen failure, the

consequence would not be significant from a-public health and safety viewpoint.

2.1.5 Independence, Redundancy and Diversity

'The staff reviewed the data -link between the safety channels and the

non-safety systems. The safety channels provide direct hard-wired scram

inputs and are also hardwired directly to independent'indicators on the

control console. The operators are provided with-information from both the

wide range and power range neutron monitors. The information is displayed on

direct analog indicators and can be displayed on a graphic CRTs. In

addition, the safety channels provide inputs to the computerized Protection,'

Control and Monitoring System (PCMS) through isolators.:-'The isolators are of

a standard commercial design. The PCMS provides connections with the

operator by controls, a keyboard. and CRT displays. -Since the PCMS does

communicate with the safety channels, this aspect of the system would not

meet the independence requirements of a power plant, but the staff concluded'

it was not necessary for the current application.

. L
- 4 -

Further, the scram circuit is essentially unchanged in that it maintains the

fail safe design using the same automatic and manual contacts which open' to

remove power to the control rod magnets. Redundant power level inputs to the

scram circuit continue to be provided. Also, additional external scram

buttons are provided for experimenters.

This system has also added the PCMS SCRAM request and PCHS watchdog scrams.

In addition to providing numerous, non-required scrams, the PCHS provides

inputs to the rod withdrawal interlocks and the reactor protection mode

interlocks. The use of'two independent power monitoring instruments and an

independent fuel temperature instrument, in addition to the watchdog scram

function provides additional diversity and redundancy to the scram system.

The system as installed meets most of the requirements of IEEE-279-1971

"Criteria for Protection Systems for Nuclear Power Generating Stationsn and

IEEE-379-1977 'Application of the Single-Failure Criteria to Nuclear Power

Generating Station Class 1E Systems."

The staff has concluded that the level of independence, redundancy and

diversity which has been maintained is acceptable.

2.1.6 Testing and Operating History

Extensive testing of the new system has been done at the licensee's facility.

Additionally, similar control systems have been used in various capacitites,

e.g., feedwater control, at nuclear power plants for sometime. The staff

also agrees with the licensee that long-term operability and safety is enhanced

due to installation of the new console system, which should provide improved

reliability, easier repair, and increased information to the operators. An

additional improvement is the self diagnostics feature which allows continuous

on-line testing and reduces the possibility of undetected failures.

2.2 Verification and Validation Plan

The staff requires an approved verification and validation (V&V) plan for

software which performs a safety function or provides information to the

operators. The software development was reviewed by the'staff to determine

the acceptability of the V&V plan. 'The staff compared the V&Y plan to

Regulatory Guide 1.152 "Criteria for Programmable Digital Computer System in

Safety Systems of Nuclear Power Generating Stations."

2.2.1 Verification of Software Design

Extensive independent verification by individuals qualified to develop such

software for the instrumentation and control console system provided acceptable

assurance that the software would function as designed. Extensive verification

testing of the software identified and resolved discrepancies and operating

problems. This verification process included consideration for breaking down

the software functions into individual discreet packages. The individual

functions of these packages were verified, in addition to the validation of

the integrated function of these packages. The staff has concluded that based

on this review, and previous experience with similar software and with the

specific vendor for this console system, the verification of the new system

software is acceptable.
- 5 -

2.2.2 Validation Testing . .

The vendor's and licensee's initial'testing program was designed to show that

the new system is capable of performing assigned tasks. 'This integrated

testing of the new console identified and resolved design problems, and

provided added assurance of proper function of.the:software. 'Additionally,ei the system has monitored reactor parameters .in parallel with.the existing..''.-

.system, and additional problems were resolved to assure acceptable system':.

function. Finally, a comprehensive process has'been devised to provide

testing and assurance that the new system willoperate properly upon.:

installation.

The staff's review finds this V&V testing.program acceptable.

2.3 Maintenance and Surveillance

An extensive program is in place to evaluate the'operation of the-components

of the control system. In.addition-to.the initial testing program,.a.-....

,self-diagnostics feature, that provides continuous on-line testing, and.routine

surveillance tasks provide additional issurance-of continued'operation'and '- rapid discovery of problem'areas. The routine surveillance tasks'include'

regularly-scheduled.surveillance and test procedures which include those items

required by the Technical Specifications as.'well as a number of other items.

Major daily checks of all scram systems, and channel tests of a'number'of

information systems are used to'document the operability of the systems.

Other tasks performed at monthly, semi-annual, and annual.intervals include

calibrations of important systems and functions.(including the controlrods,.

the power levels, the-fuel temperature and the radiation-sensing instruments)

as well as a number of non-safety-related parameters. These procedures, which

have been in place for a number of years, assured proper maintenan'ce-of the

present control system, and can be expected to provide continued assurance of

proper operation and detection of.problem areas with the proposed system.

Further, the.licensee has established'a maintenance plan'.usingthe.

manufacturer's equipment'manual'and off-line testing'package. Based on-this

experience, as well as the selfdiagnostic features of the new console system, the staff concluded that.the maintenance and testing program is acceptable for,"

the new console.

2.4 Training'of Operators

Reactor operations personnel have participated in training sessions and will.-''.

participate in formal operational testing to complete training. Operations

personnel also manipulated the controls and performed.several operations on

the new console system before installation. The staff concluded that'this

training was acceptable to assure operator understanding of-the new

instrumentation and control system.

2.5 Technical Specifications

Although the licensee provided arguments against the need for software

watchdog scram Technical Specification requirements, the licensee and the staff

recognize the desirablility of the watchdog scram. Therefore, the licensee
- 6 -

provided changes to implement the watchdog scram (Technical Specification pages

7, 16, 17, 31 and 32). The staff has concluded that these changes are needed

to acceptably assure. that- potential undetected software problems would not

result in a failure to scram or in inaccurate information' to the operator.

This requirement is consistent with the intent of. previous amendments thit

allowed installation of computerized control console systems on TRIGA reactors,

including General Atomic and the Armed Forces Radiobiological Research Institute.

Therefore, the proposed changes are acceptable.

The licensee has also proposed changes to page 14 to incorporated reactivity

addition limits applicable to the new consoles operational modes (Page 15 is

also changed to account for the retyping changes made to page 14). Based on

the staff's review of the licensee's analysis as described in the bases for

this Technical Specification change, the staff finds these changes acceptable.

3.0 ENVIRONMENTAL CONSIDERATION

This amendment involves changes in the installation or use of facility components

located within the restricted area as defined in 10 CFR Part 20 and changes in

inspection and surveillance requirements. The staff has'determined that the

amendment involves no significant increase in the amounts, and no significant

change in the types, of any effluents that may be released offsite, and there

is no significant increase in individual or cumulative occupational radiation

exposure. Accordingly, this amendment meets the eligibility criteria for

categorical exclusion set forth in 10 CFR 51.22(c)(9).. Pursuant to 10 CFR

51.22(b), no environmental impact statement or environmental assessment needs

be prepared in connection with the issuance of this amendment.

4.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the

probability or consequences of accidents previously evaluated, or create the

possibility of a new or-different kind of accident from any accident previously

evaluated, and does not involve a significant reduction in a margin of safety,

the amendment does not involve a significant hazards consideration, (2) there

is reasonable assurance that the health and safety of the public will not be

endangered by the proposed activities, and (3) such activities will be conducted

in compliance with the Commission's regulations and the issuance of this

amendment will not be inimical to the common defense and security or the

health and safety of the public.

Principal Contributor: Marvin M. Mendonca

Dated: August 6, 1991
Appendix 7.4

NRC Generic Letter 95-02

Use of NUMARC/EPRI Report TR-102348, "Guideline on

Licensing Digital Upgrades," in Determining the

Acceptability of Performing Analog-to-Digital

Replacements Under 10 CFR 50.59

STANDARD FORMAT AND CONtour
UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555

April 26, 1995

NRC GENERIC LETTER 95-02: USE OF NUMARC/EPRI:REPORT.TR-102348,: GUIDELINE ON

LICENSING DIGITAL UPGRADES;" IN DETERMINING-THE

ACCEPTABILITY OF PERFORMING ANALOG-TO-DIGITAL-''

REPLACEMENTS UNDER 10 CFR 50.59

Addressees -

All holders of operating licenses or-construction permits for-nuclear'p6wer

reactors.

Purpose . .

The U.S. Nuclear.Regulatory Commission (NRC) staff is issuing this generic

letter to inform addressees of a new staff position on the use of Nuclear

Management and Resources Council/Electrical Power Research Institute

(NUMARC/EPRI) Report TR-102348, 'Guideline on Licensing Digital Upgrades',"

dated December 1993, as acceptable guidance for determining when an analog-todigital replacement can be performed without prior NRC'staff approval under

the requirements of Section 50.59 of Title 10 of the Code of Federal;:: - Regulations (10 CFR 50.59). The report applies to all digital equipment that

uses software and, in particular,-to-microprocessor-based systems.- The .

report,.together with the clarifications discussed in this generic letter,''

represents a method acceptable.to the staff for use in making a determination

of whether or not an unreviewed safety question exists with respect to ' 10 CFR 50.59 requirements. It is expected that recipients will consider the

information in this generic'letter-when performing analog-tb-digital

instrumentation and control systems replacements. However, suggestions '''

contained in this generic letter are not NRC requirements; therefore, no

specific action or-written response is required.

Descriotion of Circumstances

The.age-related degradation of some earlier analog electronic systems and'the

difficulties in obtaining qualified replacement components for those systems,

as well as a desire for enhancedifeatures such as'automatic self-test'and

diagnostics, greater flexibility; and increased data availability have

prompted.some operating reactor licensees.to replace existing analog systems

with digital-systems., After reviewing anumber of these digital system

replacements and digital equipment failures'in both nuclear and non-nuclear

applications, the staff has.identified-potentially safety-significant-concerns

pertaining to digital systems in nuclear power plants. The concerns of'the

staff stem from.the design-characteristics specific to the'new~digital

electronics that could result in failure modes'and system malfunctions that

either.were not considered-during the initial plant design or'may not have

been evaluated in sufficient detail-in the safety analysis report.' These

9504140227
GL 95-02

April 26, 1995

Page 2 of 5

concerns include potential common mode failures due to (1) the use of common

software in redundant channels, (2) increased sensitivity to the effects of

electromagnetic interference, (3) the improper use and control of equipment

used to control and modify software and hardware configurations, (4) the

effect that some digital designs have on diverse trip functions, (5) improper

system integration, and (6) inappropriate commercial dedication of digital

electronics.

As a result of the above concerns, the NRC staff issued a draft generic letter

for public comment in the Federal Register (57FR36680) on August 14, 1992,

wherein a position was established that essentially all safety-related digital

replacements result in an unreviewed safety question because of the

possibility of the creation of a different type of malfunction than those

evaluated previously in the safety analysis report. The staff concluded,

therefore, that prior approval by the NRC staff of all safety-related digital

modifications was necessary. However, subsequent discussions and comments on

the draft generic letter have resulted in the staff position as described in

this letter.

Discussion

To assist licensees in effectively implementing digital replacements by

addressing the concerns indicated above and in determining which upgrades can

be performed under 10 CFR 50.59 without prior NRC staff approval, Report TR102348 has been published. The NRC staff reviewed and provided comments on

this report while it was in draft form, and the final report reflects a

coordinated effort between industry and the NRC staff. The-NRC staff believes

that, when properly implemented, modern digital systems offer the potential

for greater system reliability and enhanced features such as automatic selftest and diagnostics, as well as greater flexibility, increased data

availability, and ease of modification.

Report TR-102348 contains guidance that will assist licensees in implementing

and licensing digital upgrades in such a manner as to minimize the potential

concerns indicated above. It describes actions to be taken in the design and

implementation process to ensure that the digital upgrade licensing and safety

issues are addressed, and ways to consider these issues when performing the

10 CFR 50.59 evaluation. It is not the intent of the report or of the NRC

staff to predispose the outcome of the 10 CFR 50.59 process, but rather to

provide a process that will assist licensees in reaching a proper conclusion

regarding the existence of an unreviewed safety question when undertaking a

digital system replacement. However, as shown in Example 5-6 of the report,

when using this document as guidance for the analysis of modifications of some

safety-significant systems such as the reactor protection system or an

engineered safety feature system, it is likely these digital modifications

will require staff review when 10 CFR 50.59 criteria are applied. Report TR102348 states in the introduction that the guidance is supplemental to and

consistent with that provided in NSAC-125, 'Guidelines for 10 CFR 50.59 Safety

Evaluations." Licensees should bear in mind that NSAC-125 has not been
GL 95-02

April 26, 1995

Page 3 of 5

endorsed by the NRC,' and therefore any use of those guidelines is advisory

only, and that' nothing in NSAC-125 can be construed-'as a modification of

10 CFR 50.59. While the guidelines of .NSAC-125 can'be useful in the

evaluation of systems, and are representative of logic used in making a

10 CFR 50.59 determination, the actual determination of whether or not an

unreviewed safety question exists must be done in accordance with

10 CFR 50.59.

" 10 CFR 50.59(a)(2)(i) and (ii) states that a proposed change, test or

experiment involves an unreviewed 'safety question if 'the probability or

consequences of an accident or malfunction previously evaluated in the safety

analysis report may increase, or if the possibility for an.accident or

malfunction of a different type',than any previously evaluated in the safety

analysis.report may be-created..'If 'during the 10 CFR 50.59'determination

there 'is uncertainty about whether.the probability or:consequences may

increase, -or whether-the possibility of a different type of accident or

malfunction may be created, the'.uncertainty should lead the licensee 'to

concludethat the probability or consequences may increase or a new type of

malfunction may be created. '.If the uncertainty .is only on-the degree of

improvement'the digital system will provide, the modification would not

involve an 'unreviewed-safety question. If, however, the uncertainty involves

whether'or not'this modification -is more or less safe'than the previous analog

system, or'if no-degree of safety has been determined,' an unreviewed safety

question-is involved...

The staff believes that two clarifications to Report TR-102348 are appropriate

as follows:

1. 10 CFR 50.59 requires determination of whether !a possibility for an - accident or malfunction'of a different type than any evaluated

previously in the'safety analysis.report maybe created.8 As a part of

this determination,',Report TR-102348 suggests 'looking for "any new types

.of. system-level failures that would result in effects not previously

* considered in the FSAR.'," (For example, see TR-102348,- Section 4.5,

.'Question 6.) .'It'is the NRC staff's position 'that the system-level

considered'.in this regard 'should be the digital.,system being installed.

.The staff believes..that this'.clarification is necessary because'

'10 CFR 50.59 does not refer to 'an accident 'or malfunction that results

in a system-level failure different from'any previously analyzed but

rather to the malfunction of the equipment important to safety being

modified. It is the change in the.facility as described in the safety -analysis.report that is' to be analyzed under 10 CFR 50.59 to determine

if it involves an unreviewed safety question, that is, the digital - equipment that replaced''the anialog equipment,.rither than the otherwise

unchanged system of which'that 'equipment is a part .is to be analyzed.

This does not mean that'all digital equipment usage will 'automatically

result in an unreviewed safety question simply as a result of the use of

software. Software failure, including common-mode failure, must be
GL 95-02

April 26, 1995

Page 4 of 5

considered during the 10 CFR 50.59 evaluation as a possible different

type of malfunction. However, if software failure cannot cause an

equipment malfunction of a different type than any previously evaluated

in the safety analysis report, then no unreviewed safety question exists

with respect to this criterion, and in the absence of other

disqualifying criteria, the replacement can be performed under

10 CFR 50.59 without prior NRC approval. For many digital system

modifications involving relatively simple systems such as discussed in

example 5-5 of NUMARC/EPRI Report TR-102348, the NRC staff believes that

a conclusion may be reached that there is no possibility that a

different type of malfunction may be created.

As an example, when installing an upgraded digital high pressure

function of the reactor trip system, it is the digital instrumentation

and control circuitry associated with the high pressure reactor trip

function that would be subject to the questions on failure modes and

effects (equipment malfunctions) identified in the report that would be

analyzed to determine involvement of an unreviewed safety question, not

the entire reactor trip system. If the entire trip system is being

replaced with a digital upgrade, then the entire replacement digital

instrumentation and control system would be subject to the failure modes

and effects analysis, not the full range of instrumentation and control

systems being actuated to respond to a transient or'accident.

2. 10 CFR 50.59 requires maintaining records that include a written safety

evaluation which provides the bases for the determination that the

change, test, or experiment does not involve an unreviewed safety

question." Section 3.1.2 of the report points out that the use of

qualitative engineering judgment is typically involved in areas that are

not readily quantifiable, such as likelihood of the failure, its

importance to the system and to the plant, and the practicality and

incremental improvements of various options available for resolving the

failure. Such judgments may be difficult to duplicate and understand at

a later time. It is the NRC staff's position that the basis for the

engineering judgment and the logic used in the determination should be

doounented to the extent practicable. This type of documentation is of

particular importance in areas where no established consensus methods

are available, such as for software reliability, or the use of

commercial-grade hardware and software where full documentation of the

design process is not available.

EPRI Report TR-102348, together with the clarifications discussed in this

generic letter, can be used as guidance by licensees in both designing analogto-digital replacements and, with respect to unreviewed safety question

determinations, determining if an analog-to-digital replacement can be

performed under 10 CFR 50.59 without prior staff approval.
GL 95-02

April 26, 1995

Page 5 of 5

This generic letter requires no specific action or written response. If you have any questions about this matter, please contact the technical contact

listed below or the appropriate Office of Nuclear Reactor Regulation project'

manager. - -- '

. e an

Associate Direct Projects Office of Nuclear eactor Regulation'

Paul J. Loeser, NRR - l - (301) 504-2825

. -Technical contact:

'Lead project manager:

Attachment:

-Robert M. Pulsifer, NRR

(301) 504-3016 ' , .,

--List 'of Recently Issued NRC Generic Letters

.:;0 . -

'''' ' ''' ' ' '

.

.

. .

: - .

.. , v .  .

, - * . .. ..
Attachment

GL 95-02

April 26, 1995

Page 1 of 1

LIST OF RECENTLY ISSUED GENERIC LETTERS

Generic

I &++ott

Date of

Issuance Issued To

89-04,

SUPP. 1

95-01

94-04

94-03

94-02

GUIDANCE ON DEVELOPING

ACCEPTABLE INSERVICE

TESTING PROGRAMS

NRC STAFF TECHNICAL POSITION ON FIRE PROTECTION

FOR FUEL CYCLE FACILITIES

VOLUNTARY REPORTING OF

ADDITIONAL OCCUPATIONAL

RADIATION EXPOSURE DATA

INTERGRANULAR STRESS

CORROSION CRACKING OF CORE

SHROUDS IN BOILING WATER

C.

LONG-TERM SOLUTIONS AND

UPGRADE OF INTERIM

OPERATING RECOMMENDATIONS

FOR THERMAL-HYDRAULIC

INSTABILITIES IN BOILING

WATER REACTORS

REMOVAL OF ACCELERATED

TESTING AND SPECIAL REPORTING REQUIREMENTS FOR

EMERGENCY DIESEL GENERATORS

04/04/95

01/26/95

09/02/94

07/22/94

07/11/94

05/31/95

ALL HOLDERS OF OLs OR

CPs FOR NUCLEAR POWER

REACTORS.

ALL CURRENT LICENSEES

& APPLICANTS FOR URANIUM

CONVERSION & FUEL

FABRICATION FACILITIES.

ALL HOLDERS OF OLs OR CPs

FOR NPRs, RADIOGRAPHY

LICENSEES, FUEL PROCESSING LICENSEES, FABRICATING & REPROCESSING

LICENSEES, MANUFACTURERS

& DISTRIBUTORS OF BYPRODUCT MAT'L, INDEPENDDENT SPENT FUEL STORAGE

INSTALLATIONS, FACILITIES

FOR LAND DISPOSAL OF LOWLEVEL WASTE, & GEOLOGIC

REPOSITORIES FOR HIGHLEVEL WASTE.

ALL HOLDERS OF OLs OR CPs

FOR BOILING WATER

REACTORS EXCEPT FOR BIG

ROCK POINT, WHICH DOES

NOT HAVE A CORE SHROUD.

ALL HOLDERS OF OLs FOR

BOILING WATER REACTORS

EXCEPT BIG ROCK POINT

ALL HOLDERS OF OLs FOR

NPRs

94-01

OL -

CP - NPR

OPERATING LICENSE

CONSTRUCTION PERMIT

NUCLEAR POWER REACTORS

'I4
8 ELECTRICAL POWER SYSTEMS

In this chapter of the SAR, the applicant should discuss and describe the electrical

power systems at a non-power reactor facility: The electrical power systems to be

descnibed here are designed to support reactor operation. .All non-power reactors

require normal electrical service. Some non-power reactors may also require

emergency electrical service to perform finctions related to reactor safety to

ensure that, given a loss of normal electric service, sufficient power will be

available for mitigating the events discussed in SAR Chapter 13, "Acident :

Analysis." The functions to be performed and the type of emergency electricial

power systems required are developed on a case-by-case basis in other chapters of

the SAR. The information in Chapter 8 should be provided under two categones:

normal and emergency electrical power systems. " -

8.1 Normal Electrical Power Systems .. '

In this section, the applicant should discuss the design bases and functional

description of the normal electrical power systems for the reactor facility. Theinformation should include the following:

* The design bases of the noiial electric power system, including how safe

reactor shutdown Will be ensured if offsite power is lost. (This discussion

should address both short- (ransiet) and long-term electrical outages.)

* The ranges of electrical power capability required, for both reactor

operation and utilization, in terms of various principal voltages, currents,

wattage, and frequencies.

* Use of a substation, either one devoted exclusively to the reactor facility or

a shared service with other activities.

* Special processing of the electrical service by such components as isolation

;-:transformers, noise lmiters, ghtning arresters, or constant voltage

transformers.

* Schematic diagrams showing the basic distribution systems and circuits.

* Design and performance specifications of principal components, including

any that are unique of not standard.

* Special routing 6r isolatiorof wiring or circuits for both reactor operations

and experimental facilities. (The description should include provisions for

 Ab.A . R_.- v.,J .8-1 - STANDARDFOMAT-AND CO1FNT
JMLCHAFrn 8

isolaiting electrical powe service from instunertion and control crut

and safety-related circuits to avoid electromagnetic interference.)

* Any deviations or exceptions from national or local electrical codes.

* Technical specifications, if required, with bases that ensure the operability

of the normal electrical service, including surveillance requirements.

The appicant should discuss the spectrum of reactor operations that require

normal electrical power, inchlding the response of the reactor to both short

(transient) and long interruptions of normal electrical service. The applicant

should also discuss how safe reactor shutdown is ensured under all operating and

accident conditions, both with and without normal electrical service available. The

applicant should discuss how routine releases are controlled and monitored and

how the uncontrolled release of radioactive material is prevented in the event that

normal electrical power service is interrupted.

8.2 Emergency Electrical Power Systems

Emergency electrical power at a non-power reactor is defined as any temporary

substitute for normal electrical service. The various functions of emergency

electrical power can include operational convenience, assurance of experiment

integrity, and performance of functions essential to reactor integrity. In this

section, the applicant should describe all uses of emergency power systems, with

emphais on the design and functions of the emergency power systems required for

reactor safety and for protecting the health and safety of the public. All non-power

ractms should be designed for reactor shutdown in the event normal electrical

power is lost. This includes the fail-safe actuation of the control rods. Some nonpower reactors may also require emergency power to maintain the shutdown

reactor in a safe condition. Some examples of uses of emergency electrical power

follow:

* Power for reactor power level monitors, recorders, and necessary safetyrelated instruments.

* Power for effluent, process, and area radiation monitors, including

recorders.

* Power for physical security control systems, information systems, or

communications. (In this section, the applicant should only mention the

existence of such emergency electrical power and should confine details to

the fbaility physical security plan.)

Placing or maintaining experimental equipment in a safe condition.

NU1-1537,PArcr 3 REV 0,2/96

NURE 137, PART I REV 0, 2/96
.

,.ECMCAL POWER SYSTh1

* -:Power for active confinement or containment engineered safety featureI

(ESF) equipment and control systems,-such as blowers, fans, or dampers,

- and beating, ventilation, and air conditioning equipment. (This is the

equipment necessary to maintain equipment and personnel habitability or-to

control concentrations or release of airborne radioactive material and to

mitigate accident consequences.)

* Power for coolant pumps or systems that remove residual heat from the

fuel.

* Power for the emergency core cooling system, including instrumentation

and control systems.

* Power for other ESF equipment, if applicable.

* Power for emergency area lighting and communication equipment.

* Power for those instrument and control systems necessary to monitor

reactor shutdown. (These could include fuel temperature, control rod

positions, or fission product monitors.)

The types of emergency electrical power systems discussed in this section should

be commensurate with the required design bases developed in other chapters of the

SAR, such as Chapter 4, "Reactor Description"; Chapter 5, 'Reactor Coolant

Systems"; Chapter 7, "Instrumentation and Control Systems"; Chapter 9,

"Auxiliary Systems"; Chapter 10, 'Experimental Facilities and Utilization";

Chapter 11, "Radiation Protection Program and Waste Management"; and

Chapter 13. These systems may range from automatic start generators to wet-cell

or dry-cell batteries. In this section, the applicant should present a detailed

functional description and circuit diagrams. In the design bases, the applicant

should discuss if non-interruptible electrical power is required in the transfer from

normal to emergency electrical service and if the transfer is manual or automated.

The design bases should also provide voltage and power requirements for the

emergency electrical power systems, the time duration over which these could be

needed, and assurance that fuel will be available for the time required. The designs

of the emergency electrical systems should provide that any use for non-safetyrelated functions could not cause loss of necessary safety-related functions. The

design discussion should show how the emergency power supply system is isolated

or protected, if necessary, from transient effects, such as power drains, short

circuits, and electromagnetic interference. If the emergency electrical power

systems are required during analyzed accidents, the designs should include this

capability. The minimum emergency electrical power functions that would be

required to protect the health and safety of the public should be included in

technical specifications based on the discussions in Chapter 8. The technical

REv.O,2/96 8-3 STANDARD FORMAT AND CONTENT

0E.0,2/96 8-3 STANDARD FORMAT AND Cx)rNTm;
CHAFrE 8

specifications should also identify the minimum equipment to be supplied by the

emergency power system, important design parameters, and surveillance and

inspection functions that ensure operability of the emergency electrical power

systems and the supplied equipment.

_UE-57 PR _ _- ~ ,29

NUREG-1537, PART I 8-4 REV. O. 2/96
9 AUXILIARY SYSTEMS I 1. . - .

In this chapter of the SAR, the applicant should discuss the awuiliary systems at the

reactor facility. 'Auxiliary systems are those systems not fully described in other

chapters of the SAR that are important to the'safe operation and shutdown of the

reactor and to the protection'of the heilth and safety of the public, the facility staf;

and the environment. The applicant should provide sufficient information for all

auxiliazy systems to support an understanding of the design and functions of the

systems, with emphasis on those aspects that could affect the reactor and its safety

features, radiation exposures,-and the control or release'of radioactive material.

For each auxiliary system, the applicant should discuss the capability'to'function as

designed without compromising reactor operation or the capability to shut down

the reactor. This capability should be shown for normal operation and reactor

accident conditions. The applicant should include the-following information for

each auxiliary system:

(1) design bases

(2) system description, including drawings and specifications of principal

components and any special materials

(3) operational analysis and safety function

(4) instrumentation and control requirements not described in Chapter 7,

'Instrumentation and Controls Systems," of the SAR

(5) required technical specifications and their bases, including testing and

surveillanceThe design, operation, and use of non-power reactors vary widely. Typical

auary systems that may be discussed in this chapter of the SAR include the

following:

* ' Heatingehtilation, and air conditioning (HVAC) systems for normal

reactor operation. (The applicant should discuss any engineered'safety

feature functions of the HVAC systems for accident conditions in

-' Chbpter 6, 'Engineered Safety Features," of the SAR) '

* Handling and storage of reactor fuel, both new and irradiated, including

tools, vaults, racks, pools, shields, casks, and preparations for shipping.

* Fire protection systems that could affect reactor safety or protection of

licensed materials.

-REV. 0, 2/96 9-I - - Smww FORMAT Corr -REV. 0, 2/96 - - :'9-1 - -- - -- - '. STANDARD FORMAT AND CONTENT
CHIrnE 9

* Communication systems, both internal and external to the facility.

* Control, storage, or use of byproduct, source, and special nuclear material

produced, used, or possessed under the reactor operating license. (The

applicant should also discuss applicable laboratory facilities designed to

handle or use byproduct materials other than radioactive waste.)

* Cover gas control and processing at certain reactors, such as a tank reactor

using heavy water neutron moderators or reflectors. (The applicant should

include features of closed primary systems designed to control radiolytic

gases, if applicable.)

* Auxiliary coolant systems for experimental facilities and other equipment

and uses that are not part of the primary coolant system described in

Chapter 5, 'Reactor Coolant Systems," of the SAR

* Demineralizer resin regeneration system.

* Control and storage of radioactive waste and reusable radioactive

components (e.g., experiments). (The applicant should describe the

systems and show how they are designed to perform the design-basis

functions derived in Chapter 10, "Experimental Facilities and Utilization,"

or Chapter 11, "Radiation Protection Program and Waste Management," of

the SAR.)

* Control of contaminated air, gas, or liquid from experimental facilities.

(The applicant should describe the systems and show how they are

designed to perform the design-basis functions derived in Chapters 10

or 1 1.)

* Compressed air or gas systems for reactor operating systems and

experiment equipment.

* Auxiliary physical protection and access control that are not part of the

fiacility physical security plan.

These examples are not intended as a complete list of auxiliary systems that may be

disussed in this chapter of the SAR. The descriptions of some auxiliary systems

may be better suited to other chapters, which should be referenced in this section.

NUREG-1537,PaRT I 9-2 REV 0, 2/96
Auximy SYSTEMS

9.1 Heating, Ventilation, and Air Conditioning Systems

All used spaces in a facility may require HVAC systems to provide acceptable

environments for personnel and equipment. in this section the applicant should

describe how temperature and humidity are controlled and discuss the bases,

including how the control function is integrated into the HVAC systems. The

applicant should address the prevention of uncontrolled releases of airborne'

radioactive effluents to the environment for normal operation. bThe discussions

should contain explanations of how airborne radioactive material from operations

and experiments is limited in occupied areas to maintain radiation exposures below

the requirements of 10 CFR Part 20 and the facility ALARA (as low as is

reasonably achievable) program guidelines. 'Controls limiting diffusion or leakage

of radioactive material to adjacent spaces should be presented. The applicant also

should disciuss how air exhaust systems or stacks are designed to reduce the

radiological impact on the unrestricted environment during normal reactor :

operations.:

Analyses of radiation exposures in Chapter, 1 should include the applicable normal

operating characteristics of the HVAC systems described in this section of the

SAR.' The interactions among airflow patterns in the reactor room, the air exhaust

stacks, and the effluent and continuous air monitors should be discussed. If the

HVAC systems also are designed to mitigate the consequences of accidents, the

engineered safety features should be noted in this section of the SAR but described

in detail in Chapter 6.

The applicant should describe instrumentation and control systems that control the

release of radioactive material (automatic and manual) in Chapters 7 and 11 of the

SAR. The information in this section of the SAR should be sufficient to support

an understanding of the safety functions of radiation sensors that initiate alarms

and automatic closures, fail-safe dampers, interlocks, and function displays during

normal reactor operations. The applicant should discuss the bases and purpose of

technical specifications that apply to the HVAC systems, including' calibrations,

testing, and surveillance.

The applicant should disciss the possible effects of malfunctions ofthe HVAC

systems on safe reactor operation or on the release of airborne radioactive material

during normal reactor operation. The radiological effects of malfunctions should

be discussed in Chapter I.

9.2 Handling and Storage of Reactor Fuel-',".

In this section the applicant should discuss the life cycle of reactor fuel from the

time it enters its jurisdiction until it is released from such jurisdiction: Foriniost'-

non-power reactors this means from arrival on site until shipment off site., The' -

RE.0,29693 . - . SrD OMA NDCNTN

REV. 0, 2/96 9-3 _ _ STANDmDFoRmATAmCoNmirr It . -
CHAPTE 9

discussions should include descriptions of the tools, cranes, racks, design features,

and administrative controls that protect new fuel from damage.

The applicant should provide analyses and discuss how subcriticality is ensured

(k~ffnot to exceed 0.90) under all conditions, except during transportation off site.

During transportation, the shipping container license is applicable. (Existing usage

with kff greater than 0.90 will be acceptable if the usage was previously reviewed

and approved by NRC.) The applicant should address the applicability and

implementation of 10 CFR 70.24, which addresses criticality monitors.

The applicant should discuss briefly the methods that ensure the prudent control of

fuel. The discussion should include a description that does not contain proprietary

or safeguards information of the physical protection of fuel against theft or

diversion in the facility physical security plan. Reference can be made to the

physical security plan, which should be treated as proprietary or safeguards

information.

The applicant should address the handling, storage, and shipment of new and

irradiated fuel. Included should be a discussion of the tools used to insert orremove fuel from the core as well as the physical and administrative methods

specified to control their use. The details should include the design of handling

tools, transfer casks, and other radiation shields; the design of storage facilities;

methods of preparing fuel for shipping; and shipping methods. Descriptions of

procedures and systems for the storage and handling of irradiated fuel should

include radiation shielding, protection from physical damage, physical control, and

sufficient cooling to prevent overheating and surface corrosion. Irradiated fuel

cooling systems and methods may be described in detail in Chapter 5 of the SAR if

they are integral to the reactor coolant system Otherwise, the discussion should

be in this section.

During storage or handling, if a loss of fuel or cladding integrity could result in the

release of fission products, the applicant should discuss the mechanisms and

analyze the consequences in Chapter 13, "Accident Analyses," of the SAR.

Detailed discussions of radiological considerations for storing and handling fuel

should be provided in Chapter 1 and should include, if applicable, the

implementation of 10 CFR 73.6(b) concerning self-protection for irradiated highly

enriched uranium fuel.

The applicant should include in its discussions of fuel handling and storage the

bases of related technical specifications, including inspections, testing, and

surveillance and applicable administrative controls and procedures.

Additional information concerning this topic can be found in American National

Standards Institute/American Nuclear Society (ANSI/ANS) 15.19-1991.

NUREG-1S37. PART I REv 0.2/96

NUREG-1537, I REV 0. 2/96
AuxnxuRY SYsmMS

9.3 Fire Protection Systems'and Programs

In this section the applicant should describe the systems and programs designed to

protect the reactor ficility from damage by fire and discuss how the facility meets

all local building and fire codes. For a new facility this could be a general

discussion of how the facility meets local fire and building codes. Documentation

from the local authority that authorizes the construction or verifies compliance

with local codes could be submitted as' part'of the discussion. NRC construction

inspectors would review design features for fire protection during facility

construction. For existing facilities requesting license renewal forwhich the

original construction documentation might be difficult to reproduce, the applicant

could submit the results of a recent fire inspection'to show compliance with local

codes. The applicant should discuss additional active and passive design featuries

required by the reactor design characteristics. Further, the discussion should

address the potential for release of radioactive material as a result of a fire. Aative

systems might include sprinkler, suppression, hand extinguisher, and detection

systems. Passive systems might include fire walls and doors, isolation, and control

of combustible materials.

The applicant should discuss how the potential release of radioactive materials as a

result of fires in'the reactor room and other applicable spaces was considered in

the design of the facility. The discussion should include the reactor and all

facilities where special nuclear material and other radioactive materials are stored

or used underfthe reactor license. 'It should include any possible effects of a fire on

safe shutdown of the reactor. The objectives of the fire protection progm should

include the following:

* preventing fires, including limiting combustible materials

* -detecting, controlling, and extingushing fies to limit consequences.

. protecting reactor systems so that a continuing fire would not prevent safe

reactor shutdown or cause an uncontrolled release of radioactive material

The applicant should disss the bases of any technical specifications, including

testing and surveillance, as they relate to the fire protection systems and programs.

The discussion should also include the relationship between fire protection plans,

operating procedures, and the facility emergency plan.

ANSI/ANS 15.17-1987 contains general information on fire protection. The

applicant may also consult National Fire Protection Association, NFPA 802, 1993

Edition. -

I R~~v. /9.. 9.5 S.NDR FOMA .... Co.n. .. ... REV. 0, 2196 9-5 STAmARD FoRmAT Am CoNTe;r
CHAPR 9

9.4 Communication Systems

The applicant should describe the communication systems that will be used at the

facility for which public disclosure is not limited by the physical security plan.

Communication systems used between the control room, the reactor room, reactor

access point or top, reactor utilities rooms, experiment areas, and all other required

areas should be described. Such systems as telephone, paging, radio, or video that

will be used to announce changes of reactor status to experimenters, summon

supervisory operators, request radiation protection assistance, and announce

emergencies should be discussed. For a complete description of communications,

the applicant should also summarize briefly in this section the communication

systems used for emergency or physical security purposes (this discussion should

not contain proprietary or safeguards information).

The discussions of communication systems should include the bases of any related

technical specifications, including testing and surveillance.

9.5 Possession and Use of Byproduct, Source, and Special

Nuclear Material

The 10 CFR Part 50 operating license applies to possession and operation of the

reactor, possession and use of byproduct material produced by the operation of the

reactor, and, to the extent authorized, the receipt, possession, and use of other

byproduct, source, or special nuclear material (material) needed for operation of

the reactor and its eeetl programs. Examples include sources for radiation

monitor calibration, depleted uranium for shielding of experiments, reactor fuel,

fssion plates for thermal columns, and fission chambers for reactor monitoring and

control.

The NRC regulatory approach is to include in the reactor license only material that

is produced by the reactor or is required to directly operate the reactor and

associated experimental facilities. Other material at a non-power reactor facility is

authorized by an NRC byproduct, source, or special nuclear materials license. If

the facility is located in an Agreement State, an Agreement State license may also

exist This other material is normally not required to operate the reactor or

associated expeimental facilities. A special case exists for material that is received

for irradiation from another licensee. If this material is to be placed into the

reactor for irradiation within 31 days of receipt, it may be possessed under the

reactor license. However, this authorization for receipt and possession must be

specifically stated in the reactor license. If more than 31 days should pass before

the material is placed into the reactor, it should be included in an NRC or

Agreement State materials license until irradiation occurs. Further information on

this subject may be found in memoranda dated March 8 and August 18, 1988,

NUREG-1537,PT 1 9.6 REV. 0,2/96

NUREG-1537 PART I 9-6 REV. O. 2/96
AtUXImAY SYsTEms

from Dennis M. Crutchfield, Director, Division of Reactor Projects - m, IV, V,

and Special Projects, to NRC regional administrators (provided as Appendices 9.1

and 9.2).

The receipt, possession, or use of materials authorized by the reactor license may

occur Mi the reactor room and contiguous operational spaces and also in laboratory

spaces for research and development purposes. Some licensees take a narrow

view, transferring material produced in the reactor to another NRC or Agreement

State license when the material is removed from the reactor pool. Others take a

broa view and allow all materials produced by the reactor or authorized by the

license to be in various locations and laboratories in the facility. Spaces could be

.used to process and package byproduct materials for shipment or could be used for

performing experinents involving the byproduct materials. A broad view of

materials and areas authorized by the 10 CFR Part 50 reactor license avoids

maintaining multiple licenses and allows, in some cases, indemnity protection for

materials in laboratories and other auxiliary spaces. The applicant should clearly

state the materials and areas of the facility requested to be authorized by the

reactor license. 'The reactor license and technical specifications also will include

regiilatory conditions that apply to the possession, management, and use of such

nmaterials, including requirements stated in 1 0CFR Parts 20, 30, 40, or 70.

The applicant should discuss in this section laboratories under the reactor license in

which reactor-licensed material will be used. This discussion should address all

five factors noted 'at the beginning of this chapter for any such auxiliary

laboratories. The applicant should specify the types and quantities of radionuclides

authorized, as well as the general types of experiments or uses. Radiological

design bases for handling radioactive materials and radioactive waste should be

derived from Chapter'l ! of the'SAR. These design bases may'apply to chemical,

fiume, and air exhaust hoods; to drains for radioactive liquids; and to radiation

shields. The discussions should show how the physical security and emergency

plans apply to the licensed spaces and possession of byproduct materials. The

applicant should discuss the bases for special operating procedures. The

administive aspects of the use of materials in these areas should be addressed in

Chapter 12, 'Conduct of Operations," of the SAR.

9.6 Cover Gas Control in Closed Primary Coolant

Systems

Some non-power reactor designs have a reactor core tank system in the primary

coolant loop that is sealed against the atmosphere. At some of these reactors,

heavy water (a20) is used as a moderator, reflector, or coolant, and the admixture

of atmospheric water vapor and loss of the heavy water must be prevented. Atothers'a primary system operates with ordinary light water at or above atmospheric

pressure. In both types of systems;, idiolytic decomposition of the water leads to

REV. 0, 2/96 - ...- ,I- I - -- 9-7 - STANDARD FORMAT AND CONTENT
-

CHAPTER 9

a hydrogen-oxygen mixture that could reach explosive concentration without

processing.

For such reactors, the applicant should discuss gas handling in closed primary

coolant systems, addressing the five factors listed at the beaning of this chapter.

The discussions should describe cover gas systems that circulate, decontaminate,

recover, store, monitor, and dispose of the gas, as well as process or recombine

radiolytic components. The design bases should define which inert gases are

acceptable to use, their impact on safe reactor operations and shutdown, the bases

for limiting concentrations of hydrogen-oxygen mixtures, and the methods for

controlling the concentrations. The discussions should include the bases of any

required technical specifications applicable to cover gas systems, including testing

and surveillance.

9.7 Other Auxiliary Systems

As noted previously, a unique set of auxiliary systems could exist at a non-power

reactor. The above examples are found at many reactor facilities; other facilities

may have additional auxiliary systems. The applicant should describe and analyze

all auxiliary systems, address the five factors listed at the beginning of this chapter,

and include the following:

* Demonstrate that the auxiliary system will function under analyzed reactor

accident conditions, if required.

* Demonstrate that the auxiliary system and any malfunction could not create

conditions or events that could cause an unanalyzed reactor accident or the

uncontrolled release of radioactive material beyond those analyzed in

Chapter 13 of the SAR.

* Demonstrate that the auxiliary system could not prevent safe reactor

shutdown.

* Provide a discussion of and the bases for any technical specifications

related to the auxiliary system, including testing and surveillance.

9.8 References

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.17, "Fire Protection Program Criteria for Research Reactors," 1987.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.19, 'Shipment and Receipt of Special Nuclear Material (SNM) by Research

Reactor Facilities," 1991.

NUREG-1537, PART I 9-8 REV. 0,2/96

NUREG- 1537, PART I 9-8 REV. O. 2196
AuxIARY SySms

National Fire Protection Association, NFPA 802, "Recommended Practice for

Nuclear Research and Production Reactors," 1993.

REV. 0, 2/96 9-9 STANDARD FORMAT AND CoNTarr
Appendix 9.1

Regulatory Responsibilities for

Byproduct Materials in Non-Power Reactors

ST~immm FoRmAT AND CommTrT
* 0 *a

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON. D C. 2055

March.8, 1988

H. EMORANDUW FOR:. Frank J. Congel,-Director'.

Di.-_.Dvision of Radiation Safety and

Safeguards, RI . 2 . . i: , - -

Douglas M. Collins,'Director

.Division of Radiation Safety and.

Safeguards, RI2

I I. I

John A. Hind, Director

Division of Radiation Safety and

Safeguards, RIIJ

Richard L. Bangart, Director - 'Division ofRa'dation.Safety and

-Safeguards, RIV

... Ross.A. Scarano, Director :

..Divisin of Radiation Safety and

Safeguards, RV.-

FROM:

SUBJECT:

I Dennis M. Crutchfield, Director

'.:Division of.Reactor ProJects'- III,'

IY, V and Special Projects

Office of Nuclear Reactor Regulation

. . I . I , .

REGULATORY RESPONSIBILITIES FOR BYPRODUCT MATERIALS

,,IN MON-POVER REACTORS

In a memorandum dated June 8, 1987, Region IV requested guidance for

determining cases where licensed material in a non-power reactor facility may

be covered by a NRC material license or an Agreement State license, rather

than the reactor license. This issue becomes important in determining

compliance and issuing notices of violation Involving licensed material in a

reactor facility. All regions were asked to coement on this issue. After

consideration of your covments, we are providing the following guidance. The

guidance has bten coordinated with IMSS, GPA, and OGC.

1. Generic guidance related to this issue is contained in Inspection

Manual Chapter 2882, Appendices 1 and 2. Normally, material within

a non-power reactor facility will generally be assumed to be

possessed by the reactor licensee, unless there is prior documentation

approved by NRC, or some other clear demonstration that the licensed

material is covered under another license.

CONTACT:

T. Michaels

NRR/PDSNP

Ext. 21102
- 2 -

2. Consistent with I1 above, 1MS1 does not normally Issue separate

licenses which authorize possession of licensed material within an

operating reactor facility. -If a reactor facility license is silent

with regard to possession of byproduct material, it should be

amended. NRC normally exercises exclusive federal Jurisdiction

within operating reactor facilities.

3. All byproduct material which is to be inserted into a reactor, or

which is removed from the reactor, must be covered by the reactor

license while the material is within the facility.

4. The facility boundaries for a non-power reactor are normally defined

by the Safety Evaluation Report or Technical Specifications. In the

absence of identifiable facility boundaries, the Regions should

establish a facility boundary with the licensee for compliance

purposes, and the boundary should be specified in TS or FSAR.

S. As indicated in Manual Chapter 2882, Appendix 2, there are

exceptions to the above guidelines, and specific cases can be

complex. Questionable cases should be referred to Headquarters for

resolution along with a proposed course of action.

Questions concerning this guidance or specific cases should be referred to

this Division for resolution. We will coordinate with N3SS, SPA, and 0GC as

appropriate.

Division of Reactor Projeos - III. IV,

V and Special Projects

Office of Nuclear Reactor Regulation
Appendix 9.2

License Condition for Byproduct Material

To Be Irradiated in a Non-Power Reactor

STANDARD FoRMAT AND CONmmN
UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON. D.C. 205

August 18, 1988

*-EMORANDUM FOR: -Stewart D. Ebneter, Director

'Division of Radiation Safety and

Safeguards,.RI

'J.'Phillip Stohr, Director'

Division of Radiation Safety and

Safeguards, R :' -

John A. 'Hind,.Director ' .

Division of Radiation Safety and-'

Safeguards, RIII

.Richard L. Bangart, Director - -. ' . MD~v~s10n ofRadlaS10n -Safety'in '

Safeguards, RIY

Ross A. Scarano, Director' -.

Division of Radiation Safety and

Safeguards, RY

SUBJECT: 'LICENSE CONDITION FOR BYPRODUCT MATERIAL TO BE IRRADIATED

IN A'NON-POWER REACTOR

At the Reactor Health Physics Counterpart Meeting of May 18-19, 1988, a

question arose as to what the appropriate license condition should be for

possession of byproduct material at non-power reactor facilities (see

£nclosure 1, item'8). 'The question was prompted by a statement in guidance

* provided to the Regions In a memorandum dated March 8, 1988 (Enclosure 2).

-The statement in enclosure 2 appears In Item 2 and reads as follows:

...If a reactor license Is silent with regard to possession of

byproduct material it shall be amended...

All non-power reactor licenses have a license condition which permits tbe

licensee to - 'possess, but not to separate such byproduct material as may

have been produced by operation of the facility.% This license condition,

however, does not adequately cover byproduct material received at the facility

which is going to be irradiated in the reactor. Enclosure 20(Memorandum,

D.M. Crutchfield to Regions, March 8, 1988), item 3 states that -

All byproduct material which Is to be inserted into a reactor,

or which is removed from the reactor. rust be covered by the

reactor license while the material is within the facility.

In order to satisfy this condition, the license condition dealing with

possession of byproduct material should be amended if a licensee receives

CONTACT:

T. Michaels, NRR/PDSNP

492-1102
-

- 2 -

byproduct material which is to be Irradiated in the reactor. The license

condition should read as follows:

Pursuant to the Act and 10 CFR Part 30, "Rules of General Applicability

to Licensing of Byproduct Haterial,* (and Part 70)*, to receive byproduct material which is to be irradiated in the reactor within 31 days of

receipt, and to possess, but not separate, such byproduct (and special

nuclear)* materials as may be produced by operation of the facility.

* Delete if Part 70 not applicable

Licensees must request an amendment to their license to include this condition

If they receive byproduct material to be irradiated by their reactor, unless

the material is covered by another license before it is inserted into the

reactor.

Violations involving byproduct material that is to be irradiated in a

non-power reactor should generally be charged against the reactor license

unless some other specific documentation has been developed by the licensee.

In this regard the statement in enclosure 2, item 3 is modified to read as

follows:

All byproduct material which is to be inserted into a reactor, should be

covered by the reactor license; byproduct material which is removed from

the reactor must be covered by the reactor license.

Dennis H. Crutchfield, Di ctor

Division of Reactor Proj s - III,

IV, V and Special Projects Office of Nuclear Reactor Regul !p on

Enclosures:

As stated
10 EXPERIMENTAL FACILITIES AND

-UTILIZATION

-In this chapter of the SAR,-the applicant should describe and discuss the

experimental facidities at the non-power reactor facility, their intended use, and the

expe eta progra. This chapter of the SAR should contain a description of

the proposed experimental program and the safety analyses for each type of.'

experimsental fity. The design, constuction, and placement of each

experimental facility should be analyzed for inherent safety questions that exist'

apart from the experments accommodated therem. The experiments should be

analyzed by using a separate experiment safety analysis methodology to'show

compliance with the technical specifcations,' primariy the associated limiting

conditions f6r operation-(LCOs) as indicated in Chapter 14, 'Technical

Specifications," of this fornat and content guide.' The applicant should provide

sufficient information to demonstrate that no proposed operations involving

experimental irradiation or beam utilization will expose reactor operations

personnel, experimenters, or the general public to 'iinacceptable radiological

consequences. In addition to the guidance in this format and content guide,

Regulatory Guide 2.2 and 2.4 (see Appendices 10.1 and 10.2) contain guidance on

technical specifications and experimental programs'that'may be useful to the

applicant in preparing the SAR'

Non-power reactors may be used for many purposes including radiation physics,

chemistry and biology studies, materials irradiation, radionuclide production, and

educational purposes.- The experimental facilities may penetrate the reactor core

or reflector ori be located near the core. Neutron or other radiation beams can be

extracted from the core region through the biological shield. -At many non-power

reactors, the expemriental facilities are integral components ofthe entire reactor.

Utility, integrity, longevity, versatility, diversity, and safety should be considered

for the experimental facilities in the same manner they'are considered for the

reactor core and its operational coniponents and systems.- -Therefore, the safety

analyses of the reactor facility should include the experimental facilities and their

interactions'with the reactor components and systemis.; If chaiiges in reactor

operating characteristics are considered, potential interactions between'the core '-

and the experimental facilities should be analyzed. -

Experimental programs and the range of experiments vary widely among nonpower reactor facilities.- Furthermore,'as the licensee and the facility users gain

experience and as technology develops, the experimental program and many of the

specific experiments may change'overthe life of the reactor. This'nakes it very

difficult and impractical for the applicant to describe specific experiments in the

SAR. The applicant should describe and analyze in'this-chapter of the SAR anid

incorporate into the facility technical specification's enveloping conditions'of

REV. 0.,2/96 , - - , - -- R~v.. 2/9 - -'10.1- .-i ST.ANDARFoRMA"T AN CoN1tMEM
CHAPTERoO

experiment attributes such as reactivity limits or material properties to allow the

greatest flexibility in the experimental program. Potential experimental needs

should be considered when establishing these limiting safety aspects in the SAR, so

that determinations in accordance with 10 CFR 50.59 can be made expeditiously.

Experience has shown that most licensees have successfully implemented changes

in experimental programs without prior NRC approval under the provisions of

10 CFR 50.59. This regulation allows licensees to (1) make changes in the facility

as described in the SAR, (2) make changes in the procedures as described in the

SAR, and (3) conduct tests or experiments not described in the SAR without prior

NRC approval, unless the proposed change, test, or experiment involves a change

in the technical specifications incorporated in the license or an unreviewed safety

question. A proposed change, test, or experiment is deemed an unreviewed safety

question if(1) the probability of occurrence or the consequences of an accident or

malfunction of equipment important to safety previously evaluated in the SAR may

be increased, (2) a possibility for an accident or malfunction of a type different

from any evaluated previously in the SAR may be created, or (3) the margin of

safety as defined in the basis for any technical specification is reduced.

Some non-power reactors are operated as critical facilities to demonstrate fuel

loading and perform reactor physics studies. In such cases, the reactor itself can

also be considered an experimental facility. In this case, the safety analysis and

experiment technical specifications will include the limitations on core

configurations and operational limitations when the core is the experiment.

The applicant should provide an analysis to demonstrate that the reactor and

experimental facilities can be operated safely. This analysis should include the

range of normal operations, accidents, and malfunctions of experimental facilities.

It should address any impact the experimental facility imposes on the reactor and

any impact the reactor imposes on the experimental facility.

Consideration should be given to the possibility of the experimental facility causing

an accident that requires analysis in Chapter 13, 'Accident Analyses," of the SAR.

In some cases, the failure of an experiment can be the maximum hypothetical

accident (MHA) for the reactor. This possibility is most prevalent with fueled

experiments. Experiments can result in the maximum uncontrolled reactivity

addition accident at a facility. Limiting experiment failure should be considered in

Chapter 13.

The SAR should be written to accommodate the nature of varying experiments and

meet the requirements of future experimentation. The applicant should show that

there is no undue risk to the health and safety of the public.

Discussions in this chapter of the SAR should include design bases, facility

descriptions, functional and safety analyses, and the applicant's safety conclusions

NUREG-1537, PART I 10-2 REV 0,2/96

NUREG-1537, PART I 10-2 REV 0, 21/96
BEXPMMENTALFAIMMES AND UTILIZATION

for all expeimenta facilities. The structural design and its potential impact on

reactoroperation should be analyzed for those aeperimental facilities that are

permanently attached to the reactor support structure, reactor vessel, or pool

hardware. For those experimental facilities that penetrate the reactor vessel below

any primary coolant water level, an analysis of the experimental design should

demonstrate that the design is resistant to failure and that if failure occurs, it is

bounded by the anialysis in Chapter 13 of the SAR for a loss-of-coolant accident

(LOCA). The placement or use of experimental facilities shall not compromise the

functionality of any reactor safety system or engineered safety feature. Thediscussion should include the capabilities, limitations, and controls on reactor

operation, including engineering or procedural controls for experiments, that

ensure radiation doses do not exceed the requirements in 10 CFR Part 20 and are

consistent with the facility program to keep exposure to radiation as low as is

reasonably achievable (ALARA). . - '

Because of the potentially unlimited variety of experiments that can be

accommodated in a non-power reactor, the applicant should show that

administrative:'controls are adequate to ensure that the health and safety of the

public are protected. The actual experiments to be performed need not be

discussed in detail in this chapter of the SAR, but the limiting and enveloping

features of the experiments and the administrative procedures used by the applicant

to review, approve, and safely control experiments should be described. The

applicant'should provide the bases for experiment-related LCOs and for a detailed

description and justification of the experiment review and acceptance program that

are then specified in the technical specifications. -

10.1 Summary Descrnption

In this section'of the SAR, the applica 'sh6ouldbriefly' describe the principal

features of the experimental and irradiafion facilities associated with the reactor.

The applicant should discuss the scope of the experimental program and define

what is considered tobe an experiment. Discussions should include experimental

compatibility iithnormal reactor operations and accidents and measures taken to

avoid interference with the reactori shutdown and other systems.,

The applicant should include the following information: '

* geneial focus of th` experimental program (radiation science, medical,

materials iesting, teaching, etc.)

a list of experimental facilities

* : basic type of experiments that will be conducted (incore, thermal column,

exteiiial beam,'etc.)

R~v.,2/6 1-3 SANDRD ORMA AN COTEN

REV. O. 2/9  .: 10-3 - - I- STANDmD FoRmAT Am CoNmrr
CHAPTER 10

* a brief description of experiment monitoring and control and the interaction

between the experiment and the reactor control and safety systems

* a brief overview of design requirements for the experiment and of the

review and approval process

Simple block diagrams and drawings may be used to show the location, basic

function, and relationship of each experimental facility to the reactor. The

summary description should contain enough information to support an overall

understanding of the functions of the experimental facilities and the experiment

review and approval process.

A brief description of typical experimental facilities found at non-power reactors

follows. This list, however, is not exhaustive.

* Incorefacilfiies are those facilities that are surrounded on at least two

sides by fuel. Such facilities are commonly called void tubes, flux traps,

central irradiation facilities, incore irradiation facilities, radioisotope

facilities, dummy and demountable fuel elements, fist and thermal neutron

irradiation facilities, or central and offset thimbles. If the cross-sectional

area of an incore facility is greater than 16 square inches, the reactor is

considered a test reactor if the thermal power level exceeds 1 megawatt.

The facility is also considered a test reactor if there is a circulating loop

through the core for conducting fuel experiments and the reactor power

exceeds 1 megawatt.

* In-reflectorfacilities are those facilities that are physically located in the

reflector and are surrounded either on all sides or on at least three sides by

reflector material. In-reflector facilities might include lazy susans, void

tubes, flux traps, thimbles, standpipes, or thermal neutron irradiation

facilities.

* Automatic trcmrferfacilities, sometimes called rabbits, are a special class

of incore and in-reflector experimental facility. They often protrude into or

are adjacent to the core or reflector and contain the experimental material.

However, rabbit facilities allow the experimental material to be moved

quickly into and out of the desired flux region of the core by pneumatic,

hydraulic, or mechanical means. The material can be moved while the

reactor is operating if limits on reactivity changes in the reactor are

observed.

* Beam ports are hollow tubes that can abut the core or protrude into the

core or reflector. However, unlike the previously described incore and inreflector facilities, they may or may not contain the experimental material.

NUREG-1537,PART I 10-4 REV 0,2196

NUREG-1537, PART I 10-4 REV 0, 2196
ExPER= ENTAL FACILES AND UTiLZATnON

Instead, they may be used to channel radiation from the core to a position,

usually outside the reactor vessel and the biological shield, where the'

experiment is located. Neutrons and gamma-ray beams are tailored to suit

the experimental needs. -

.hermal columns function in a way similar to that of beam ports in that

they allow transport of radiation away from the core to areas where the

experiment is located. Rather than a tube to guide radiation beams, they

consist of a neutron moderator, typically a large volume of graphite blocks,

*enclosed in a container. The column is located at one face of the reactor in

: place ofthe reflector. Fast neutrons are thermalized within the moderator.

and may be used outside or inside the reactor shield for experiments.

*; . Irradicon rooms or other dry cavities in the biological shield may be

located adjacent to the reactor core (or the core moved into position) for

irradiation of large volumes of material or objects.,

* Cold neuhron source is a special type of beam port. The neutrons are

passed through a very cold moderator, such as frozen heavywater or,

hydrogen cooled by active cryostatic systems, to reduce their energy to

below the normal thermal range and increase the relative flux densities of

very slow neutrons. This allows a wider range of materials to be probed

and the probability of interactions in some materials to increase. These

cold neutron beams are sometimes used with neutron guides, which can

carry the neutrons substantial distances from the reactor without significant

* losses. . ..

If the source includes hydrogenous neutron moderators and cryogenic:

equipment, unique safety questions could arise. Such a source could be

included in the initial reactor design, obtained by later modification of an

existing reactor, or installed in an approved experimental facility, such as a

beam port orthermal column, of an existing reactor..-^,-

10.2. Experimental Facilities.. -

In this section of the SAR, the applicant should describe and discuss in detail all

experimental facilities. The design should ensure that risks to the public, stafi and

experimenters are acceptable. . ....1 :.- ,..C. .- r . . I ' : C' "''

The applicant should discuss specifications and important design and operatingparameters for the experimental facilities and give design details and the physical .

size, including all dimensions. Simplified engineering drawings or schematics may

be used, especially for more complex facilities. The applicant should discuss the'

location of the experimental facility in relation to the core, safety systems, core

REv.O,2196 - 10-5 - STANDARD FORMAT AND CONTENT

REV. 0, 2/96 . ..- 10-5 -1 STAMAM FORMAT AND CONTENT
CHAPTER IO

support, neutron detectors, coolant system components, and any other reactor

systems, components, or structures.

Features of the experimental facility that could interfere with safe reactor

shutdown or with adequate core cooling shall be included. The source of

experiment cooling and any dependence on or interaction with the reactor coolant

system should be discussed. For any experimental facilities that require a special

cooling system independent of the reactor primary coolant system, the technical

evaluation considerations are similar to those for the reactor coolant system. The

applicant should follow the applicable guidance in Chapter 5, 'Reactor Coolant

Systems," in this format and content guide for independent experiment cooling

systems.

Since integrity of the experimental facility is important, the capability to contain or

withstand any postulated pressure pulse and preclude any inadvertent primary

coolant leakage or facility collapse should be discussed. Analysis should be

presented for vessel or pool penetrations that could affect the risk of a LOCA. For

experimental facilities that penetrate the reactor vessel below the water level of the

pool surface, the applicant should show that if a LOCA does occur, the

consequences are bounded by the LOCA analysis in Chapter 13 of the SARI The

LOCA analysis should also apply to the experimental facility if the facility must be

cooled.

The applicant should discuss the materials used in the construction of the

experimental facilities, addressing radiation and chemistry impacts. Materials and

design, including physical dimensions, should limit any rapid reactivity insertion if

the facility is suddenly voided or flooded. The supporting analysis should be

included in Chapter 13 of the SAR, where change in the limiting experiment failure

reactivity should be analyzed. The bases of applicable LCOs for the technical

specifications should be developed and justified.

The radiological considerations associated with the design and use of the

experimental facilities, generation of radioactive gases (including argon-41),

release of fission products or other radioactive contaminants, and exposure of

personnel to neutron and gamma beams should be summarized in this section of

the SAR and discussed in greater detail in Chapter 11, "Radiation Protection

Program and Waste Management."

Direct radiation streaming from the experimental facilities and the effect of

scattered (skyshine) radiation should be discussed briefly in this section of the SAR

and analyzed in Chapter 11. The analysis should clearly show all pertinent

radiation sources, distances, dimensions, materials, radiation scattering, and

material attenuation factors.

NUREG-1537, PART I 10-6 REv 0,2/96

NUREG-1537,PRT I 10-6 REV 0, 2196
ExRENTAL FAcIrriEs ASD UnuzATnoN

Facilities that could fail and release argon-41 or other airborne radioactivity into

the facility air or to the environment should be analyzed. The analysis in

Chapter 13 of the SAR and summarized in this section should show the

concentrations of radioactive material in the experimental facility, the release

pathway, and the concentrations of radioactive material in the reactor facility and

the outside environment. In some cases, this type of failure could be the MHA for

the reactor, which should be analyzed in Chapter 13.

Any radiation monitors specifically designed and placed to detect expriment

radiation and to monitor personnel should be discussed briefly in this section of the

SAR and discussed in greater detail in Chapters 7, "Instrumentation and Control

Systems," and 11. Additionally, reactor operating characteristics, including scrams

and runbacks associated with experimental measurements, should be analyzed.

Any physical restraints, shields, or beam catchers, both temporary and permanently

installed, that are used to restrict access to radiation areas associated with

experimental facilities should be described and analyzed. Descriptions and

analyses should show that the placement, dimensions, and materials (1) are

sufficient to limit the expected radiation doses to experimenters, reactor operators,

and other personnel to levels below those specified in 10 CFR Part 20 and (2) are

consistent with the facility ALARA program. For reactor beams, the applicant

should describe the approach to compliance with the regulations concerning access

to high radiation areas and very high radiation areas, as appropriate. These issues

should be analyzed in Chapter 11 and sumnarized in this section of the SAR,.

Permanentiy installed safety instrumentation for the experiment facility, including

the location and function of sensors, readout devices, and scram or interlock

capabilities, should be discussed briefly in this section of the SAR and in greater

detail in Chapter 7. , - ;

In addition to the applicable information discussed above, this section should

contain the following for cold sources because of their unique safety'

considerations' ' : -

* description of the cold source facility, including the operating principles

and the design of the systems and components -

description of the relevant ambient environmental conditions, such as

radiation'intensities and actual theimal sources, and their potential imnpact

on the cold source components and miaterials'

* discussion of the physical and chemical characieristics of the neutron

moderator and coolant fluids, handling systems, volumes and state's of

REV 0,2/96 -10-7 STANDARD FOIMAT CoitNT  1, Rn, 0, 2/96 . I. --10-7 STAwDARD FoRmAT AND CoNrnaT
XUCHARM 10

matter at the operating temperatures and at ambient temperature, and all

hardware, shielding, control, and safety features of the cold source

description of all applicable operations for preparing and using the facility,

such as inserting and removing moderator and coolant fluids; storing,

sensing, and measuring inventories; determining contamination and

leakage; determining chemical and physical changes of fluids and

interactions with hardware; and producing and monitoring the operating

temperatures and pressures

* discussion of the effect of the radiation environment, such as radiolysis and

other radiolytic changes, on fluids, formation or release of ozone and other

gas, heating of fluids and components caused by radiation and by

conduction or convection from nearby shield and structural components,

and radioactivity of fluids and components,

* discussion of the effect of leakage of fluids, such as toxicity, flammability,

and potential to detonate, addressing changes in composition, mixtures, or

other characteristics of the neutron moderator and coolant fluids with use

and cycling

* description of provisions for safe (passive) shutdown of the cold source

and reactor as a system

Technical specifications for experimental facilities, as discussed in Chapter 14 of

this format and content guide, should be presented and justified in this section of

the SAR

10.3 Experiment Review

Because of the variety of experiments that can be conducted in a non-power

reactor, the administrative controls of the applicant should be adequate to ensure

the protection of the public. The administrative procedures used by the applicant

to review and approve experiments should be described in detail in this section of

the SAR and summarized in Chapter 12, 'Conduct of Operations," and the

operating limits should be included in the technical specifications. The applicant

should state the safety analysis requirements for the experiment safety analysis

report and the experiment review and approval methodology and should briefly

discuss the authority and role of the experiment review committee.

The applicant should discuss experiment classification and approval authority. The

applicant should state the methodology used to categorize proposed experiments

according to risk potential, the categories expected at the reactor facility, and the

safety requirements for each category. The methodology should describe how

NUREG-1537,PART I 10-8 REV. O. 2/96
ExPrERIEAL FAcILTIEs AND UTnLzAnoN

10 CFR 50.59 will be used in the review of all experiments not described in the

SAR, as well as how Regulatory Guides 2.2 and 2.4 (see Appendices 10.1 and

10.2) will be used. The appropriate level of review'authority required to approve

experiments in each category should be discussed. The applicant should be

specific in delineating the bounds of the risk categories, such as gram amounts,

temperature degree limits, radioactivity limits, or reactivity limits, and should

develop the bases of applicable technical specifications. The experiment safety

analysis process should demonstrate compliance with these limits and establish any

special controls on the experiment.

The applicant should discuss administrative controls for the experiment and list the

administrative controls used to protect facility personnel and the public from

radiation or other possible hazards, such as chemical releases, in the

implementation of the experimental program. Where appropriate, the discussion

should delineate areas where reactor operations and experiment operations are

performed under separate authority and by different personnel. The discussion

should include access to experimental facilities and areas, lockout procedures,

communications with reactor operating personnel, alarms, and reactor scrams.

The administrative procedures should address basic protection and recovery

procedures after a malfunction of experiments or experimental facilities.

The applicant should discuss the generic safety assessment of experimental

materials and limitations, consistent with the guidance in Regulatory Guide 2.2,

from which experiment and reactor LCOs are incorporated in the technical

specifications. Malfunctions or failures of experiments with significant potential

for radiological consequences should be analyzed in Chapter 13 of the SAR and

summarzed in this section. For some reactors, the most serious accident or the

MHA could be initiated by an experiment malfunction. Areas of assessment

should include the following:

* fissile materials and radiological risks from radiation fields or release of

radioactive material

* trace elements and impurities

* effects on reactivity, both positive and negative

*' explosive, corrosive, and highly reactive chemicals

* radiation-sensitive materials

* flammable or toxic materials

* cryogenic liquids

.

REV. 0, 2/96 -. 10-9 STANDARD FORMAT AND CONTENT

' '- REV,. 02/96 :,,.,~ 10-9 STANDARD FORMAT AND CoNnm4T
CHAPTER IO

* unknown materials

* radiation heating or damage that could cause experiment malfunction

* heating that could cause departure from nucleate boiling on surfaces

<2,

-- - _ . . 0. 2/96

NUTREG- 1537, PART I 10-10 REV. O. 2/96
Appendix 10.1

Regulatory Guide 2.2

Development of Technical Specifications for

Experiments in Research Reactors

STANJDARD FoRMAT Am Co'Nmrr
Nowamber 1973 U.S. ATOMIC ENERGY COMMISSION

REGULATORY GUIDE

OIRECIORATE OF aEGULATORY STANDARD

REGULATORY GUIDE 2.2

DEVELOPMENT OF TECHNICAL SPECIFICATIONS FOR

EXPERIMENTS IN RESEARCH REACTORS

A. INTRODUCTION

Paragraph SO.34(bX4) -of 10 CFR Part 50. . Licensuig of .Production and Utilization Facilities,"

requires thit each application for an operating license

provide a final analysis and evaluation of the design and

performance of structures systems, and components of

-the facility with the objective of assesing the risk to

public health and safety resulting from operation of the

facility. Section 50.36 of 10 CFR Part 50 requires that

each such apphation also include proposed technical

specifications derived from the analyses and evaluation

perforned for the safety analysis report (SAR)

This guide describes information that should be

included in proposed technical specifications for

experiments in research reactors It identifies

considerations that should be addressed in the evaluation

of experimental programs as well as corsiderations that

should be* addressed. to define limnts and other

requirements to be included in the technical

specifications. It is expected that the guidelines

delineated here will be adapted, as required, to specific

features ard characteristics of mdividual research

reactors.

- 1. DISCUSSION

Each safety analysis report (SAR) contains a

description of the proposed experimental programand

safety analyses for each type of experimental facility

proposed. It includes descriptions of and safety analyses

for permanently mistalled facihties such as beam tubes,

thermal columns, hydraulic or pneumatic tube systems,

and other types of capsule Trradiation facities, and

movable experimental facilities (in some types of.

reactors) which accommodate placement of shells; tubes,

trays, baskets, or other guidng or positionung devices in

'or adjacent to the reactor, core. Safety analyses for

special modes of reactor system or component use to

accommodate individual, repetitive, or mult;ple

expenments should also be prov-ded. These can inclode

such categories as reactor pulsiag, use of reactor coolant

or fuel as 'gmma radiation sources. or use offuel in

subcritical arrays separated from the core.

' The design, construction,' and placement of each

experimental facility should -beanalyzed for inherent

safety -questions that exist apart from experinents

accommodated therein. In addition,' for each

experimental facility and mode of reactor system or

component use. the descnptiorns and safety analyses

should address the types and scopes of experiments

intended to be performed.

The purposes of presenting tuch'sifety analyses ac

(I) to demonstrate that the experimental program a,

envinsoned at the time of presentrtion of the SAR can be

carried out without undue isk to the public health and

safety, (2) to demonstate the technical ability to carry

out the kind of safety analyses which is expected to be

done on a continuing basis throughout the evolution of

the experimental program, (3) to establish bases apinst

which unreviewed safety questons can be measured

pursuant to paragraph (c) of §5059, and (4) to develop

subject matter approprate for mclusion in technical

specifications

Safety ktresearch reactor experimentation requwres

that consideration be given to ary feature of the design

or conduct of an experiment, including intended

functions and poswble malfunctions, which can create,

directly or mdirectly, a radiological exposure haurd

Safety analyses for experiments should consider (I) any

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inicrAr-tion of in experiment with the reactor system

tIlut has the potential for breaching any primary barrser

to fission product release from fuel. (2) any interaction

-II in expenment with the reactor system that could

.Jvenbely iffect any engineered safety features or

,ontrol iystem features designed to protect the public

Iorni i lasnin product release, (3) any inherent feature

cii ili experiment that could create beams, radiation

iicIdw. or unconfined radioactive materials, and (4) any

potentially adverse interaction with concurrent

experimental and operational activities.

A variety of specific technical factors, considered

igainst the foregoing cnteria can give nse to safety

problems as follows

I Factors in experiments which could cause a breach

in any of the fission product bamers.

a. Reactivity effects as a result of placement or

removal of an experiment or of motion of material

wiltiun the experiments due, for example to forced or

natural convection of fluwds, phase changes, chemical or

ridiolytic dissociation, or mechanical instability.

b Thermal effects on fuel which alter local heat

generation oreheat transfer rates as a result of neutron

flux perturbations, gamma heating, electrical heating, or

alteration of coolant temperature or flow by experiment

comnprents or failure thereof due to heating, radiation

degradation, or radiolytic dissoeation.

c Mechanical forces on fuel cladding arsing from

the manipulation of experimental components, from

tools used for such manipulation, from thermal stress,

vibration, or shock waves, or from missiles arising from

tunctioning or malfunctioning expernments.

d. Chemical attack, including corrosion, resulting

from the use in or escape of materials into the fuel

environment or accelerated corrosion due to elevated

tmperatures.

: Factors in experiments which could adversely affect

engineered safety features or control system features.

A Neutron flux perturbations affecting

calibrat ions of safety channels and/or rod worths

b. Mechanucal forces adversely affecting shielding

or wonfinement ansing from causes as in Ic. above.

* c. Radiation fields or radioactive relcses from

expcrnients which can mask the performance of an

uOperational monitoring system intended for the

detection of fission product releases at early stages.

d Physical interference by experiment

t.mponents with reactor system components such as

*ontrol or safety rods or physical displacement of

reajor system shielding.

. Fai~tors in experments which could create

radiological nsks due to radiation fields or unconfired

rJclditL1lve material

J Uge of materials which are or become

,licmiul ly unstable or highly reactive or are subject to

buildup of temperature or pressure. eg., preusure

buildup in special beam port plugs.

b. Irradiation of finely divided solids, liquids. or

pie which are readly airborne if inadequately

confined.

c. Degradation or failure of ma:erials intended to

confine experiments, e.g., by radiation decomposition of

nonmesallic captules, weld faiures, gasket failures,

excessive internal heat generation, or inadequate cooling.

d. Depadation or failure of vent systems or filter

installations or inadequate uhelding thereof.

e. Degradation or failure of safety-related

mnstruments or control devices on experiments

f. Mechanical Instability resulting in unintended

movement of an experiment relative to its shielding, eg.

by faulty scking of lead brick, by exceeding floor

loading capablities, or by capsules becoming buoyant in

water.

* g. Use of inadequate devices for suelding and

handling experiment components or capsules following

Irradrations.

4. Factors relating to interactions with other

experiments or with operational activities.

a. Reactivity effects of concurrent motion

occurnng in two or more experiments.

b. Potentially adverse interactions resulting from

the use of common ectinc circutts and supplies and

common portions of fluid systems such as manifolds for

cooling water, vent, or drain systerrs.

c. Physical interference by experinents with

patterns of operational activty which could Impede or

prevent a sfety or emergency function, ep.g blocking of

access routes.

d. Creation ofi mdustrial hazards such as the

generation or release of toxic at noxious mateals which

could rnpenr the ability of operators to perform

necessary reactor safety functions.

e Special modes of reactor operation such as

pulsing abnormal occurrences in reactor operation or

reactor accidents which could tr2Ega failures in

expetiments.

The proposed techncal specfications that are

relevant to experiments in research reactors should (I)

have bases relating to safety considerations as requred

by §SO6a). (2) address subject areas that are clearly

under the direct control of the licensee, and (3) faIl

under the categories of limiting conditions for operation.

surveillance requirements, des gn features, or

administrative controls, as specified in §5036

Situations may anse in which the safety analyses of

some unique experiments establish the need to colwider

the effects of such experiments on the safety lirmts and

limiting safety system sertuigs for reactor operation

Tecbhucal specifications should provide reasonable

flexibility to perform experiments, install new

experimental facilities, or change or remove from use

.2-2

'KJ'

K>
taiLltics previously described. Proposed technical

specifications should address' safety-onented

considerations, as distinct from functional or end-use

dwicriptions of experimental programs. On the other

lund, all safety considerations implicit Zn each individual

experiment proposed must be enumerated and evaluated

to determine whether or not they fall witlun the safety

inalysis for reactor operation presented mn the SAR. In

iddition the proposed experiment should be evaluated in

detail and its execution controlled so as to reduce any

ladiation dose to plant personnel and the public to the

lowest practicable level.

C. REGULATORY POSITION

The safety-oriented considerations from which

Icilinical specifications for expertments should be

developed include (I) the physical conditions of the

design and conduct of experiments, (2) the materials

content of expeiments, and (3) the adsiunistratie

controls employed to evaluate, authorize, and carry out

cxpenments The material that follows is organized

according to thecabove three considerations, but it is not

intended that this be the only tormat acceptable tor use

lor proposed technical specifications. The definitions of

certain terms used in ths section are given in Appendix

A

1. Physical Conditions

a Reactivity Effects

From a safety standpoint, the princapal concerm

Is that associted with a net positive reactivity effect,

whether it is caused by the insertion of an expeinment

havinga positive reactivity effect or by the removal of

an experiment having a negative reactvity effect. Credit

may be taken for the operation of the reactor safety

system and engineered safeguirdt systems provided (I)

they have been designed to standards and critera

establishing very high reliabty, such as ASI N42.7

(IEEE.279), (2) adequate quality asurnCe was provided

in their construction aand is provided durisn operation,

and (3) it can be shown that they can function

independently of the assumed experiment falure mode.

All proposed transients should be analyzed to assure that

a safety limit would not be exceeed.

(I) Every experiment should be evaluated for

its static reactivity worth and its potential reactivity

wortlL

(2) The potential reactinty worth of each

-ecured renovable experiment should be less thin that

value of reactivity which. If introduced as a positive step

change, could result in a transient that would be likely

to lead to doses ra any restricted or unrestncted irea in

excess of the limits set forth xn 10 CFR Part 20.

(3) The magnitude ot the potential reactivity

worth of each unsecured experiment should be less than

that value which, if introduced au a posuitie step change

in reactivity, would cause a viola-ion i1 .sfely binil or

'of the mnnmurn shutdown magtr.

(4) The rate of changc ot reactivity ol any

unsecured expenment, any movaale experiment, or any

combination of such expenrients introduced by

intentionally setting the experiment(s) in motion relative

to' the 'ieactor should not exceed the capacity of the

control system to provide compensation.

(S) The sum of the magnitudes of the static

reactvity worthis of all unsecured expernments which

coexist should not exceed th! maximum value or

potential reactivity worth authorized for a single secured

removable experiment or 'the minimum shutdown

margin, whichever as less.

b. Themsal-Hydraulic Effects

- (I) Every expenrment should be evaluated for

its actual and potential thermal effects or reactor

components 'and coolant. Normally, ths evaluation

should be made'fot the reactor at the extremes of its

operating' margin.'as defined by linmting safety system

settings. '

(2) Expernments should be designed to prevent

the negation of any'flux pekinug or reactor coolant flow

considerations 'that have been used to define or are

implicit in the safety limits for the reactor Coolant flow

considerations should include potential blockage or

redistrdbution sand potential phase changes in liqud

coolant.

(3) The surface temperature of the material

wuhch bounds'ot supports any experiment should not

'exceed the lowest of the foUowing where apppbcable ' ' I ' (a) 'the saturation temperature of liquid

reactor coolant at any point of mutual contact. ' ' (b) a' temperature 'conservatively below

that at which the corrosson rate' of the boundary

material tt sny suface would lead to its fadure;' or,

(c) a temperature conservatively below

'th.at at which the strength of the boundary material

would be reduced to a point predictably leiding to

failwe. .'

c. Mechatsiel StRM Effects

E()very experiment should be evaluated with

respect 'to' the storage and possble uncontrolled rdease

of any mechanical energy.

(2) Experiments mvohing a potential for

creating' objects with substantial momentum (nussiles)

shoiuld be oriented in such a way as to numnize the

probability of damage to the reactor systern

(3) Materials of construction and fabrication

and iassembly techniques utilized in experiments should

be so specified and used that assurance is provided that

no stress faflure can occur at stresses twice 'those

antimpated tn the manipulation and conduct of the

experiment or twice those which could occur as a result

of unintended but credible changes of, or within, the

expnesient.

22.3
(4) Prototype testing under experiment

Londitions should be employed to demonstrate the.

ihbility lo withstand failure

2. Material Content of Experiments

Certain kinds of naterials which may be used in

experiments possess properties with sgnificant safety

implitcJtions Limitations on the amounts of such

materials can limit the consequences, of experiment

failures The material content of every experiment

should be analyzed and limuted according to the

classrfications given below.

a. Radioactive materials

(I) The radioactive material content, including

fission products, of any singly encapsulated experiment

should be limited so that the complete release of all

gaseous, particulate, or volatile components from the

encapsulation will not result in doses mn excessof 10% of

the equivalent annual doses stated in 10 CFR Part 20

This dose bimit applies to persons occupying (I)

unrestricted areas continuously for two hours starting at

time of release or (2) restricted areas during the length

-of time required to evacuate the restricted area.

(2) The radioactive material content, including

fission products, of any doubly encapsulated or vented

experiment should be limited so that the complete

release of all gaseous, psrticulate, or volatile components

from the encapsulation or confining boundary of the

experiment could not result in (I) a dose to any person

occupying an unrestricted area continuously for a period

of two hours starting at the tune of release in excess of

0.5 rem to the whole body or I1S rem to the thyroid or

(2) a dose to any person occupying a restncted area

during the length of time required to evacuate the

restricted area m excess of S rem to the whole body or

30 rert to the thyroid.

(3) For purposes- of applying the above

considerations, a single-mode nonviolent failure of the

encapsulation boundary that releases all radioactive

material into the immediate environment of the

experiment or to the reactor building, as appropriate,

should be assumed. The analysis should establish the

most probable traiectory of the material, if any, into

restricted and unrestricted are*s Credit for natural

consequence-limitrig features such as solubibty.

absorption, and dilution and for installed features such

as filters may be taken provided each such feature is

specifically identified and consesvatively justified by

specific test or physical data or well~itabbshed physical

mechanisms. In addition, with 'respect to installed

features.,oredit taken for their effectiveness should

depend a the adequacy of the related quality assurance

prooctures undertaken, including the extent to 'which

surveillance test% simulate the conditions to be met in

practice If assumptions regarding atmospheric dilut)on

*re involved. they should not be less conservative than

those used in the analysis of Design Basis Accidents

Irradiation of riksionjble materal<

excluding the fissionable material content of fuel

element assemblies described in the technical

specifications, should be deemed an unreviewed wfetv

question unless a specification meeting the above criterw

and its related safety analysis have been approved by the

Commission. With respect to other radioactive materials.

specifications and safety snalyses should be submitted

that are representative of experiments vrth either the

highest inventory of radioactive material; or the highest

probability for failure that could result in the escape of

such materul :nto restricted and unrest-3cted areas. In

addition, recoris should be generated and maintained to

allow for review to demonstrate that the radioactive

material content of each individual expe; iment does not

exceed that allowed by the stated criteria.

These considerations should not be

tnterpreted (I) to permit or encourage any unnecessary

intentional releases of radioactive materials to

unrestricted areas, or (2) to reliee the obligation ta

mm rze and control radiation doses in Restricted areas

b. Trace Elements and Imperities

A reasonable effort should be made to identify

in advance of an expenrment trace elements or impurities

whose act:vation products may represent the dominant

radiological hazard.

c. Higl-CrosScectioi Materials

Nuclides possessing high thcrrroal neutron

absorption cross - sections should be identified and

linited with respect to their quantity or method of

inclusion in individual experiments in order to control

reactivity or tiermal effects within the hmitations

d. Htigly Rcactive Ciealkah,

The inclusion of explosuve matersals in

experiments constitutes an unretewed safety question

unless such tsage has been reiewed and approved b) the

Comnrnison, except that amounts up to 23 milrlams of

tIM equivalent may be irradiated or stored inside the

reactor confinement system in accordance with

regulatory position C.lc.

e* Coerwie Cbei"

A lit should be prepaed identifying materials

which are demically incompatible with the reactor

system from the viewpoint of corrosion and which

should be excluded from any experiments or the use of

which is sukect to special scrutiny and control. This list

should be provided to all who use the reactor.

f. Radiatiou-Sertive Materials

The evaluation of each experiment shoulu

include an assessment of the consequences of physical or

I. . ..

22.4
chcinku.l change, in the material content as a result Qf its

presence in 2 radiation environment, particularly for

tionnietdllic materils

hffe ts lo be considered include the alterution

ot degridiaion of mechanical properties due to

ijdition-miduced decomiposition, eg., of plastic; or

polymncr., and radjolytlc generation of excessive pgs

premurc or explosive gas mixtures.

g. Fammable or Tpxic Materials

Procedures control should incorporate

mechanisms for handling and limiting the quantitis of

highly flammable or toxic materials used in experumgntal

programs or used in the reactor room.

h. Cryogenic Uouia

The inclusion of cryogenic liquids within the

biological shield of a research reactor would constptute

mn unreviewed safety quesuon unless such usage has

been reviewed and approved by the Commission.

i. Unknown Materuals

No experiments should be performed unless the

itiJter 1 content, with the exception of trace

c.nmtituents, is known.

3. Administrative Controls of Experiments

a. Internal Authoijation

(I) Evaluation ty Safety ReiewGroup

(a) No experiment should be performed

without review and approval by a technically competent

S.ifety Review Group or Committee. RepetItive

experiments with safety considerations in common pay

be reviewed and approved as a class.

(b) Criteria for review of an experiment or

JlAss of expenments should include (1) applicble

regulatory cnteria, including those in 10 CFR Par 20

dnd the technical secifications and (2) in-house safety criteria and rules which have been established for facpity

upcritionu. including those which govern requirenwnts

lor eiiaptulation. venting, filtratIon, shielding, and

%amibr experiment design considerations, as well as those

which govern the quality assurance program required

under § 50.34.

(0) Records should be kept of the Safety

Review Group's review and authorization fo: each

experiment or class of experiments.

(2) Operatios Approval

(a) Every experiment should have the

prior explicit written approval of the Licensed Senior

Operator in charge of reactor operations.

(b) Every person who is to carry out an

experiment should be certified by the Licensed Senior

Operator in charge of reactor operations as to the

sufficiency of his knowledge and training in procedures

required for the safe conduct of the experiment

b. Procedres for Active Conduct of Experimnts

(I) Detailed written procedures should be

provided for the use or operation of each experimental

facility.

(2) The Licensed Operator at the console

should be notified just prior to moving any experiment

within the reactor area and should authorize such

mowement.

(3) Each experiment removed from the reactor

or reactor system should be subject to a radiation

monitoring procedure which anticipates exposure rates

greater than those predicted. The results of such

monitoring should be documented.

c. Prpeedues Reting to Personnel Accs to

Experiments

(I) There should be a documented procedure

for the control of visitor acess to the reactor area to

inimize the likelihood of unnecessary exposure to

radiation as a result of experimental activities and to

minimize the possibility of mttentuonal or unintentional

ob truction of safety.

(2) There should be a wnren training

procedure for the purpose of qulifyg expernmenters in

the reactor and safety-related aspects of their activities.

including their expected responses to alarms.

d. Quality Amurnnce Proguim

There should be a Quality Assurance Program

covering the design, fabrication, and testing of

experiments, including procedures for verification of

kinds and amounts of their material contents such as

those described in regulatory position C 2.

2.2-5
APPENDIX A

DEFIN ITIONS

1. Expeiment--An experiment, as used herein, is any

of the following

a. An actrvity utiling the reactor system or as

components or the neutrons or radiation

generated therein,

b. An evaluation or test of a reactor system

operational, surveillance, or maintenance

technique;

c. An experimental or testing activity which as

conducted within the confimement or

containment system of the reactor, or

d. The material content of any of the foregong.

including structural components, encaptulation

or confining boundaries, and contained fluids

or solids.

2. Experimental Facility-An experimental facility is

any structure or device which is intended to guide,

orient, position, manipulate, or otherwise facilitate

a multiplicity of experiments of similar character.

3. Exploaive Material-Explonve material is any solid

or liquid which is categorized as a Severe,

Dangerous, or Very Dangerous Explosion Hazard in

'Dangerous Properties of Industrial Materials" by

N. 1. Sax, Third Ed. (1968), or is grven an

Identification of Reactivity (Stability) index of 2, 3,

or 4 by the National Fire Protection Association in

us publication 704.M 1966,ZIdentificatmon System

lor Fire Hazards of Materuls," also enumerated in

the "Handbook for Laboratory Safety" 2nd Ed.

(1971) published by The Chemical Rubber Co.

4. Movable Expement-A movable experiment is one

which may be inserted, removed, or manipulated

while the reactor is critical.

S. Potential Reactivity Worth-The potential reactivity

worth of an expernment is the maximum absolute

value of the reactivity change that would occur as a

result of intended or anticipated changes or credible

malfunctions that altet experhment position or

configuration.

The eviluation must consider possible trajectories of

the experiment in motion relative to the reactor. lts

orientation along each trajectory. and circumstances

which can cause internal changes such as creating or

filling of void spaces or motion of inecdical

components. For removable experiments. the

potential reactivity worth ts equal to or greater than

the static reactivity worth.

6. Removanle Experiment-A removable experiment is

any experiment, experimental facility, or

component of an experiment, other than a

permanently attached appurtenance to the reactor

system, which can reasonably be anticipated to be

moved one or more times during the life of the

reactor.

7. Secured Experiment-Any experiment,

experimental facility, or component of an

experiment is deemed to be secured, or in a secured

position, if it is held in a stationary position relative

to the reactor by mechanical means. The restraning

forces must be substantially greater than those to

which the experiment might be subjected by

hydraulic. pneumatic, buoyant, or other forces

which are normal to the operating environment of

the experunent, or by forces which can arise as a

reslt of credible Malfunctions.

8. Static Reactivity Worth-As used herein, the static

reactivity worth of an experiment as the absolute

value of the reactirvty change which is measurable

by calibrated control or regulating rod comparison

methods between two defined terminal positions or

configurations of the experiment. For removable

experiments, the terminl positions are fully

removed from the reactor and fully inserted or

nistalled in the normal functioning or ntmded

porition.

9. Unsecured Experiment-Any experiment,

experimental facility, or component of an

experiment is deemed to be unsecured if it as not

and when it is not secured as defined in 7. above..

dovug parts of experiments are deemed to be

unsecured when they ar in mntion.

2.2-7
Appendix 10.2

Regulatory Guide 2.4

Review of Experiments for Research Reactors

STANDARD FORmAT AND CoNzTNT
U.S. NUCLEAR REGULATORY COMMISSION July 19 76

REGULATORY GUIDE

OFFICE OF STANDARDS DEVELOPMENT

REGULATORY GUIDE 2.4

REVIEW OF EXPERIMENTS FOR RESEARCH REACTORS

*. - .A. fINTRODUCTION

Section 50.36, ."Technical Specifications," of 10 CFR Part 50, -

"Licensing of Production and Utilization Facilities," requires that each

applicant for -a license authorizing operation of a production or utilization facility include in his application proposed technical-specifications. If.acceptable, these technical specifications, along with any

other such specifications that.the Coumission finds appropriate, are incorporated into the facilitylicense that is issued by the Commission

and are conditions of the license.

Paragraph (c)(5), "Administrative controls," of 5 50.36 of 10 CFR

Part 50 requires that technical specifications for nuclear reactors include provisions relating to the organization and management procedures,

recordkeeping, review and.audit, and reporting necessary to ensure

operation of the facility in.a-safe manner. Section 50.59, "Changes, - tests and experiments," of 10 CFR Part 50 permits each holder of i license

authorizing operation of a production or utilization facility tobtake

changes in the .facility and procedures as described in the safety analysis

report (SAR) and to conduct tests or experiments not described in the SAR,

without prior Commission approval,.unless the proposed change, test, or

experimentlinvolves a change in the technical specification incorporated

in the license or an unreviewed safety, question.

This guide describes procedures acceptable to the NRC staff for the

licensee's reviev and approval of experiments performed at research reactor

facilities.

B. DISCUSSION

Standard ANSI N401-1974 (ANS-15.6), "Review of Experiments for

Research Reactors,"* was prepared by Work Group ANS-15.6 and sponsored by

* . . . -. ; . o . . ,. . .. .

CopsLes may be obtained from the American Nuclear Society,.244 East Ogden

Avenue, Hinsdale,.Illinois 60521.

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Subcommittee ANS-15 (Research Reactors) of the American Nuclear Society

(ANS). This standard was approved by the American National Standards

Committee N17, Research Reactors, Reactor Physics and Radiation Shielding,

and its Secretariat in March 1974. It was subsequently approved and

designated ANSI N401-1974 by the American National Standards Institute

(ANSI) on November 19, 1974. The standard provides guidance for the

licensee's review and-approval of experiments performed at research reactor

facilities by identifying substantive areas for each experiment that should

be reviewed to provide assurance that the experiment (1) falls within the

limits delineated in the technical specifications, (2) does not present an

unreviewed safety question as defined in §50.59, "Changes, tests and

experiments," of 10 CFR Part 50, (3) does not constitute a threat to the

health and safety of any individuals, and (4) does not constitute a hazard

to the reactor facility or other equipment. In addition, this standard

recommends a system for classifying experiments to establish levels of

licensee review and approval commensurate with the level of risk inherent

in the experiment. Both the requirements and the recommendations of the

standard have been evaluated by the staff in evaluating the acceptability

of this standard.

C. REGULATORY POSITION

The requirements and recomcendations provided in ANSI N401-1974,

"Review of Experiments for Research Reactors," are-generally acceptable to

the NRC staff. The guidance provides an adequate basis for the review and

approval of research reactor experiments performed in accordance with

5§50.36 and 50.59 of 10 CFR Part 50, subject to the following:

1. The last sentence of the paragraph defining, "shall, should and

may," as given in Section 3, "Definitions," of ANSI N401-1974 should be

modified to read as follows: "To conform to this standard, experiment

review shall be performed in accordance with the standard's requirements

and recommendations."

2. The definition of non-secured experiment, as given in Section 3,

"Definitions," should be modified to read as follows: "Any exseriment,

experimental facility, or component of an experiment is considered to be

unsecured when it is not secured as defined under secured experiment in

Section 3."

3. Subsection 4.1, "Classification System," should be modified by

adding the following sentence: "The experiment classification system to

determine level of approval for the experiments should be reviewed and

approved by the Reactor Safety Committee designated in the Technical

Specifications."

4. In addition to the experiment plan (Section 5, "The Experimental

Plan"), there should also exist detailed procedures for carrying out an

2.4-2
experiment, and these procedures should be reviewed as required by the

facility's technical specifications. A single experimental procedure may

be used for more than one exposure or more than one identical experiment,

but such a procedure should expire after a specified interval.

5. Subsection 6.1, "Review Proce'dure," should be modified by adding

the following sentence: "The experiment review procedure should be

reviewed and approved by the Reactor Safety Committee designated in the

Technical Specifications."

6. Paragraph (3) of Subsection 6.2, "Considerations," should be

replaced with the following: "Does the experiment meet all criteria

regarding reactivity effects? These criteria include assurance that (1)

the potential reactivity worth of each secured experiment would be less

than that value of reactivity which, if introduced as a positive step

change, could result in a transient that would be likely to lead to doses

in any restricted or unrestricted area in excess of the limits set forth

in 10 CFR Part 20; (2) the magnitude of the potential reactivity worth of

each non-secured experiment would be less than that value which, if introduced as a positive step change in reactivity, would cause a violation of

a safety limit or of the minimum shutdown margin; and (3) the rate of

change and magnitude of reactivity of any moveable experiment, moveable

parts of experiments, or any combination of such experiments introduced

by intentionally setting the experiments in motion relative to the reactor

would not exceed the capacity of the control system to provide

compensation."

7. In Subsection 6.3, "Review Personnel," the last paragraph should

be replaced with the following: "Members of the Committee should disqualify themselves from the review of experiments in which they are

directly involved. They may act as consultants to the review group but

should not be involved with the final decision for approval or disapproval

of the experiment."

8. The specific applicability or acceptability of items 1, 3, 4, and

5 of Section 9, "References," of ANSI N401-1974 will be covered separately

in other regulatory guides, where appropriate.

D. IMPLEMENTATION

The purpose of this section is to provide guidance to, applicants

regarding the NRC staff's plans for using this regulatory guide in the

review of research reactor facility applications. The staff will use this

guide in evaluating applications submitted after the date shown below.

However, all or part of this guide may be used by the staff to the extent

reasonable and practicable for evaluating prior applications. Such use is

"cually reflected in the staff review questions and subsequent evaluations

2.4-3
for specific cases. Backfitting action, if required, will be considered

separately pursuant to Section 50.109 of 10 CFR Part 50.

Except in those cases in which the applicant proposes an acceptable

alternative method for complying with specified portions of the Commission's

regulations, the method described herein will be used by the staff in

evaluating applications in connection with research reactor facility construction permits, operating licenses, or proposed amendments thereto

submitted for approval after March 1, 1977.

U!NITEO STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON. , 0 C POSTAGE AND FEES PAID

OFFICIAL BUSINSS OCCULATOOV COC u.1ata

PtEAITY FOR PRIVATE USE. X00i

2.4-4
11 RADIATION PROTECTION PROGRAM AND

WASTE MANAGEMENT

In ths chapter of the SAR, the applicant'should discuss and analyze all radiological

consequences related to normal operation of the reactor. In general,'the design

and fmction of stnucturessystems,'and co'mponents arnd al facility'opeations and

mhaerials aathorized by ithe'operating license should be desiribed in detail in other

chapters ofthe'SAR. Chapter 11 should co the a p ducieal oiscussions of the

ficility program to control radiation and expected exposures due'tooperation,

maintenance, and use of the reactor. In this chapter, the applicant ihould develop

the methods for quantitative assessment of radiation doses in the restricted,

controlled (if present), and unrestricted areas; should apply those methods to all

applicable radiation sources related to the full range of operation; should describe

the'program and provisions for protecting the health and safety of the public

(including workers) and the environment; and should provide the bases for

analyzing radiological consequences from potential accidents addressed in detail in

Chapter 13, "Accident Analyses."'

In accordance with 10 CFR 20.1 101, it is the responsibility of the applicant to

develop, document,'and implement a radiation protection program' conmnensurate

with the scope and extent of licensed activities and sufficient to'ensure compliance

with the regulations'in 10 CFR Part 20: To the extent practicable,Ithe applicant

should also use procedures and engineering controls based on sound principles of

radiation protection to keep doses to occupationalvworkers and members of the

public as low as is reasonably achievable (ALARA).

Waste materials resulting from maintenance, normal operations, or accident

conditions at non-power reactors may contain radioactive isotopes. Such wastes

are governed by the operating license, and, like other licensed materials, they must

be controlled. At a non-power reactor, management and control responsibility for

radioactive waste may be assigned tothe organization responsible for reactor''

operations, and the radiation protection organization may provide independent

oversight for monitoring, assessing, and limiting riski related to radiati6n sources.

Alternatively, facility management could assign primary responsibility for handling

and disposing of radioactive wastes to the organization responsible for radiation

protection. In either case, the applicant should require procedures to ensure that

radiation exposures and releases of radioactive material are adequately assessed

and controlled. :The applicant should discuss these issues and submit the

information necessary for NRC review. This forxnat and content guide for

Chapter 1, of the SAR integrates radioactive waste management and radiological

protection in some sections, and provides separate sections for some information.

The applicant should organize the functions and present the information as best

suits the facility consistent with this guide.

REV. 0,2196 11-I STANDARD FORMAT AND CorENT

REV. 0,2/96 11-1 STANDARD FORMAT AND CONTENT
CHATR 1I 

11.1 Radiation Protection

The sections that follow provide guidance on the information the applicant should

include in the description of the radiation protection program. The program is

applied to the design of the reactor and its equipment, the experimental facilities,

reactor operations, design and use of associated laboratories, planning and

procedures, and the instrumentation, techniques, and practices employed to verifyr

compliance with the radiation dose limits and other applicable requirements

specified in the regulations. Plans and the bases used to develop procedures for

assessing and controlling radioactive wastes and the ALARA program should be

included. The responsibilities of the health physics organization at the reactor

facility, as well as any other applicant radiation protection organizations (e.g.,

under a separate materials license), should be described. Facility organization

charts should be included that show independence of the radiation protection

function from the facility operations function.

In this chapter, the applicant should address all radiation sources and radioactive

materials produced in the reactor and possessed or used within the reactor facility

under the authorization of the reactor license. Other byproduct, special nuclear

material (SNM), and source material possessed or used under the authorization of

the reactor license but not produced by reactor operation should be described.

Program details should be given in the sections that follow.

11.1.1 Radiation Sources

In this section, the applicant should describe the sources of radiation that are

monitored and controlled by the radiation protection and radioactive waste

programs. In general, the sources should be categorized as airborne liquid, or

solid as discussed in the sections that follow.

The applicant should include in this description a tabulation of all standard, check,

and startup sources categorized by isotopic composition, principal radiations (e g.,

beta and gamma ray energies, abundance > 10%), activity (curie content), neutron

characteristics, geometry, physical and chemical form, and whether the source is

sealed or unsealed.

The applicant should also tabulate all fissile and fissionable materials, including fuel

elements and assemblies, showing the status (fresh, in-core, interim storage, or

spent), original enrichment [including uranium-235 (U-235) and total uranium (U)

content], and current enrichment [including current U-235, total U, and total

plutonium (Pu) (if appropriate)].

Because of the varied nature of experimental programs, the source strengths of

irradiated experimental materials are not necessarily tabulated in an SAR.

NUREG-1537, PART I 11-2 REV 0,2196 NUREG- 1537, PART I 1-2 REV 0, 2/96
RADIATnON PRtaECnON PROGRAM AND WASTE MANAGEMENr

However, the full range of source strengths expected to be encountered in the

experimental program should be listed and discussed. Experimental protocols

should provide detailed source data and should be subject to the review of facility

operations staff, the health physicist, and, in the'case of new experjmnents and'

specified deviations from previous experiments, the reactor review or audit

committee. In evaluating all experiments, the applicant shall also consider the

requirements'of 10 CFR 50.59.

Conservative estimates should be made ofthe quantities and types of radioactive

wastes expected to result fromrreactor operations and use, based on'previous or

other similar reactor facility experience. Identification of such wastes should

distinguish, if possible, which are associated with the operation of the reactor and

which are associated with the utilization ofthe reactor, if utilization occurs under

the reactor license. Non-power reactor applicants have 'a'tendency to provide

.overly conservative estimates; although estimates should be conservative, they

'hould also be realistic. tu s s b conevti

Where feasible, the applicant should include the physical and chemical form,

amounts, use, storage conditions, and locations of all sources: In occupied or

accessible areas, conservative estimates of external radiation fields should be given.

An estimate of the maximum annual dose and collective doses to workers and the

public should be given for majo'r and repetitive activities involving radiation: 'The

applicant should discuss hlow the requirements of Subpart C of 10 CFR Part 20

(20.1201-20.1208),'which contains regulations for occuipational dose limits, and

Subpart D of 10 CFR Part 20 (20.1301-20.1302), which contains regulations for

radiation dose limits for individual members of the public, will be met. Regulations

concerning compliance with dose limits for individual mermibers of the publicare

given in 10 CFR 20.1302. Applicants that have licensed non-power reactors

usually have historical information on radiation doses. They should discuss this

'information.'

License c6nditions and, if applicable, technicil specifications, concerning material

possession limits, enrichment, material formsand source strengths, should be

developed'and analyzed in this and other chapters, such as 'Chapter 4, "Reactor

Description," of the SAR These will control the use of the sources discussed

above. -'

11.1.1.1 Airborne Radiation Sources

-'Airborne' radioactive sources should be described in a manner suitable for'

designing worker protective'measures and assessing and controlling workers'

doses. Airborne radionuclides' are important because'they typically are the

principal source of radiation exposure to the public from'a'non-power reactor. In

a table, the applicant should summariie'the predicted concentrations and quantities

REV. 0./9 .1- Sm1 ,FRA N OTN

'REV. O. 2/9 11-3 . STANARD FoRmAT AND CoNTENT
CHAPTER 1I

of airborne radionuclides during the full range of normal operation (which includes

maintenance activities) according to the areas that could be occupied by personnel.

The applicant should estimate the release of airborne radionuclides to the

environment and should use these releases to determine consequences in the offsite

environment. The applicant should discuss compliance with the applicable

regulations (10 CFR Part 20). Note that while airborne radioactive sources from

accidents are discussed in Chapter 13, the calculational methodologies developed

here should be applicable to accident release analysis. Therefore, the models and

assumptions used for the prediction and calculation of the dose rates and

accumulative doses in both the restricted, controlled (f present), and unrestricted

areas should be provided in detail. The guidance that follows gives an example of

a description of appropriate methodology as illustrated for argonl-4 (Ar-41) that is

also applicable to any airborne radionuclide, provided both internal and external

dose delivery are accounted for.

The potential for argon-41 production exists at most non-power reactor facilities

over the full range of normal operations, and argon-41 could be the predominant

radionuclide released to the unrestricted area. Argon-41 is produced when the Ar40 in air and air in solution in water is activated by neutrons. Argon-41 may be

considered a radioactive waste produced by reactor operations. The specific

source locations (e.g., primary coolant water, beam tubes, exposure rooms, and

air-driven rabbit systems), predicted production rates, release mechanisms and

rates, concentrations in occupied areas, possible personnel doses and dose rates,

release points from the restricted area, dilution air (quantities and sources),

quantities and concentrations predicted to be released, annual average atmospheric

conditions, diffusion and dispersion, predicted concentrations in unrestricted areas,

and potential dose rates and annual doses, including gamma-ray shine from

elevated plumes, should be addressed in detail.

For argon-41, as well as other noble gases at non-power reactors, it is acceptable

to assume that all significant radiation risk is from external exposure to beta and

gamma radiation. Other radionuclides (e.g., halogens or particulates) could cause

internal radiation risk by being ingested or inhaled. All these doses should be

addressed, as applicable. The assumptions and methods should be conservative

but physically realistic, and the validity of dose calculations should be assessed.

Some non-power reactor applicants have used conservative assumptions and

methods that have resulted in answers that, although acceptable, are conservative

by large factors. The applicant should consider discussing the amount of

conservatism built into the calculations. All assumptions should be justified, and

sources of information should be adequately referenced. The calculations should

address possible doses in the restricted areas, in the controlled areas (if applicable),

and in the unrestricted areas. In the unrestricted area, potential doses should be

analyzed for the maximally exposed individual, at the location of the nearest

permanent residence, and at any locations of special interest, such as a classroom

NUREG-1537, PART I 11-4 REV O,2J96 NUREG 1 537, PMT I 114 REV 0, 2196
RDIA:TION PROTECnON PROGRAmANDWASTEMANAGEMENT

or a campus dormitory. Due care should be taken if finite or non-uniform airborne

distributions are intermingled with infinite cloud approximations within buildings

or in idealized gaussian plumes. Any such intermingling of models or assumptions

should bejjustified. Similar discussions in this paragraph of the SAR should

address the production of airborne particulates, aerosols, vapors, and nitrogen-16

(N-16) or other radionuclides.-

The discussion and calculations should show how the facility design ensures that

doses to the facility staff and the public will not exceed '10 'CFR Part 20 limits and

that its ALARA requirements for effluents are satisfied.

11.1.1.2 'Liquid Radioactive Sources

The applicant should identify all expected liquid radioactive sources, such as

reactor primary coolant, experimental solutions, reference sources, and fissile

material.' The applicant should identify their origin and should specify whether

they result from reactor operations or the utilization program or whether they exist

for special purposes. Information should include radionuclides, concentrations,

total curie strength, solubility, container characteristics, and planied release or

disposition. Liquid radioactive wastes should be included. However, since the

types of such wastes, their origins, and the source strengths will vary with time and

with the nature of the utilization program, only limited descriptions of liquid

wastes should be provided. The'applicant should estimate the quantity of liquid

effluent released to the unrestricted envir nmedn. The applicant should discuss if

credit is taken for dilution preceding release. The applicant should discuss

compliance with the applicable sections of 10 CFR Part 20, such as.10 CFR

20.2003 and any disposal of licensed materiWl approved under 10 CFR 20.2002.

Any storage or disposal facilities should be noted, with reference to their

management and use and the design basis of their radiation protection capabilities.

11.1.13 Solid Radioactive Sources

The applicant should identify all expected solid radioactive sources, such as reactor

fuel (spent, in-core, and fresh), calibration and test sources, experiment samples,

and facility components. The information should include, among other things,

radionuclides, curie strengths, and physical characteristics and whether the source

is sealed or unsealed. Solid radioactive waste should be noted, but because the

types and quantities will vary with time and the utilization program, 'only limited

descriptions of solid wastes need b' provided.' Provisions for classifying,

monitoring, storing,- packaging, volume reduction prior to shipment, and disposing

of solid radioactive wastes should be discussed. The applicant should estimate the

annual volume of solid waste expected to be removed from the site 'and its

radioactive content (in curies). The applicant should discuss compliance with

REV. 0, 2/96 0/ 15 .ON

I I . . 4

STANDARD FoRmAT AND CoN7ENr
CHATER I I <-'

applicable sections of 10 CFR Parts 20, 61, and 71, and Department of

Transportation regulations (49 CFR) for transporting radioactive material.

The applicant should discuss any capabilities or approvals received under NRC or

State material licenses for onsite or offsite storage of solid radioactive wastes,

including how the necessary characteristics of a restricted area are maintained.

The applicant should discuss any disposal of licensed material approved under

10 CFR 20.2002.

This section should contain the design bases for temporary, permanent, and

installed shielding components at the facility, including utilization, laboratory

facilities, and radiation beams.

The following areas of the facility should be examined when developing the

program fbr inventory and control of radiation sources:

* the exterior of the reactor biological shielding and reactor auxiliary

locations (e.g.. primary coolant system components and dernineralizers)

accessible to personnel

* the reactor experimental facilities, including beam ports, thermal columns,

pneumatic or hydraulic transfer facilities, and all other irradiation facilities

* the radioactive material handling, preparation, packaging, and utilization

facilities, including laboratories, hot cells, caves, and storage and

processing areas

* other extraneous sources, including, for example, neutron and gamma

irradiation facilities, check and standard sources, neutron sources, fuel

handling and storage facilities, experimental equipment storage facilities,

and radioactive waste handling and storage facilities

11.1.2 Radiation Protection Program

In this section, the applicant should describe the structure of the organization that

administers the radiation protection program required by 10 CFR 20.1101,

including information about staffing levels, positions of authority and

responsibility, and position qualifications. Working relationships with other safety

organizations, including the reactor facility operations staff, should be described.

The applicant should discuss the charters, standards, procedures, and other

documents that specify the authority and responsibilities of the organization,

including authority to interdict perceived unsafe practices. The administrative

plans and procedures that implement the facility policy, the overall program, and

the way the organization, policy, and program are designed for effective operation

NURBG-1537, PART I 11-6 REV 0.2196

NUREG-1537,PART ) 11-6 REV 0, 2/96
should be discussed. In this discussion, the applicant should describe the

management policy governing the program and the allocation of policymalding

responsibilities. Reference can be made to Chapter 12, "Conduct of Operations,"

if such information appears there. -

The information should include the document control measures employed to

ensure that the plans and procedures relative to the radiation protection program,

including changes, are reviewed for adequacy, approved by authorized personnel,

and distributed to and used by the applicable staff at the locations where radiation

exposures could be encountered.

The radiation safety training program should be described in detail. This

discussion should give the scope, and a summary of the content, of the training

provided or required for all personnel, including facility-employed personnel

health physics personnel, non-facility-employed research and service personnel,

visitors, and security, fire, and other emergency personnel.

The applicant should describe the purpose, organization, and functions of any

review and audit committees with responsibilities relating to radiation safety,'

including the'charter, frequency of meetings, audits, scope of any reviews, and

qualifications and requirements for committee members. The applicant should,

describe how each committee's work relates to the radiation safety organization

and how a comprehensive program is ensured. If this information is discussed in

Chapter 12, it can be referenced here.

The program for conducting ficility radiation safety audits of all functional

elements of the radiation protection program to meet the requirements of 10 CFR

20.1 101(c) should be described, iderffifying the scope of the audits, the bases for

'scheduling the audits, the qualifications of the auditors, the management level to

which reports are sent, and the process for following upion audit findings. The

relationship of this program to any other self-assessmentlinterinl appraisal

program should be discussed. 'The bases for technical specifications related to

facility radiation safety audits should be provided.

The system that examines the experiences of the radiation protection program and

uses these experiences to improve the program and the facility desig'foriradiation

protection should be described. -This system should also examine problem and

incidents and develop lessons 'learned," root causes, and effective corrective

actions. . - , , t .. ._s . ' ,

For activities not described in the SAR or not governed by procedures, a work

control process such as the use'ofradiation work permits should be used.' The

applicant should discuss the control program used at the facility.

REV.~~~ 117_~~wFRA _.19 ~ OT~

PRFV. 0. 2/96 11-7' STANDARD FoRmAT AND CoNTENT
CHAPTER II

The applicant should describe the recordkeeping process for the radiation safety

program, including record-retention periods, accessibility, review, and archiving.

Review of radiation safety records for accuracy and validity should be discussed.

The use of records for developing trend analyses, informing management, planning

radiation-related actions, and reporting to regulatory and other duly authorized

entities should be discussed.

11.13 ALARA Program

In this section, the applicant should describe the ALARA program for the facility

required by 10 CFR 20.1101. The description should include the basis for the

program and the management level and authority by which the facility ALARA

policy is established. The applicant should discuss how this program is

implemented to maintain radiological doses of all personnel at the facility and

releases of effluents to the unrestricted area ALARA. The applicant should discuss

the criteria used to determine how low the projected doses should be to permit

task implementation (i.e., ALARA goals). The discussion should include methods

to ensure that the radiation protection staff with their considerations of the facility

ALARA program, are specifically involved during review and approval of design,

in construction of facilities, in the planning and implementing of reactor utilization

(experiment design and planning) and operation, in maintenance activities, and in

the management and disposition of radioactive wastes.

11.1.4 Radiation Monitoring and Surveying

The program employed to routinely monitor workplaces and other locations

accessible to people for identification and control of sources of radiation exposure

should be described in this section, including the measures designed to ensure that

air, liquids, and solids are monitored in all applicable areas. The applicant should

also discuss the bases of the methods and procedures used for detecting and

assessing contaminated areas, materials, and components, and should describe the

records that document the applicability, quality, and accuracy of monitoring

methods, techniques, and procedures.

The applicant should provide summary descriptions of all radiation monitoring

equipment employed throughout the facility, including locations and functions of

each device and system. This summary should also describe sampling equipment

for liquid and gaseous process and effluent streams. This discussion may be

combined with (and appropriately cross-referenced to) the discussions in Cbapter

7, 'Instrumentation and Control Systems." The applicant should discuss the

interface between the radiation monitoring system and engineered safety features

discussed in Chapter 6, 'Engineered Safety Features," if any exist. Types of

equipment should include systems of the following types (as appropriate to the

facility):

_U _AR E-57 _ _1- _~ 0.2

NUREG-1537, PART I I1-8 REV 0, 2/96
RADIAnON PRoTECTloN PROGRAM AND WAsTE MANAaEmENr

* cotinous ir onitors (CA?&), including fixed and moving filter, and

gaseous monitors

* portable survey instruments (radiation fields and contamination)

-remote area monitors (RAMs)

* -samplers -

* effluent monitors

* environmental monitors (details should appear in Section 111.7)

* personal dosimneters

* portal monitors,

- radwaste storage monitors -

* -criticality monitors

The calibration of radiation protection instrumentation, including the procedures

and standards governing calibration, control of the calibration process, use of

national standards, and verification should be described. In this section, the

applicant should also describe the calibration equipment and discuss sensitivities to

environmental and other conditions with respect to the calibration requirements.

The program to ensure that routine periodic calibration is performed in a timely

manner and the bases of calibration schedules should be described.

The applicant should describe how routine monitoring provided at the ficility is'

planned to'ensure that radiation exposures to the public and workers or material

releases can be detected, and should discuss how the approach used for routine

monitoring provides reasonable assurance that all radiation at, and released from,

the site'will be appropriately monitored. -'

Technical specifications and their bases related to the radiation monitoring

equipment and procedures, as discussed in Chapter 14, 'Technical Specifications,"

should be justified in'this section.

11.1.5 Radiation Exposure Control and Dosimetry

Radiation 'exposure is controlled by controlling radioactive materials and effluent

radioactive material releases. In this section of the SAR, the applicant should -

describe'the design bases for the equipment and procedures utilized for controlling

REv. 0, 2/96 -, .11-9 STANDARD FORMAT AND CONENTr
CHAPTERII

exposures to personnel and releases of radioactive materials from the facility, and

should discuss how the facility structures, systems, and components are designed

to provide assurance that there will be no uncontrolled effluent radioactive releases

to the environment or to work areas. Some systems, such as containment,

confinement, and ventilation, may have been discussed in other chapters of the

SAR; reference to those discussions in this chapter of the SAR is appropriate. The

applicant should also discuss how the bases of radiation shielding, ventilation, and

remote handling and decontamination equipment are designed to ensure that doses

to the workers are maintained ALARA and within the applicable regulatory limits.

How the design of required entry control devices (i.e., alarms, signals, or locked

entry ways) alerts workers to, or prevents entry into, high radiation and very high

radiation areas should be described. The regulations in 10 CFR Part 20, Subpart

G. Control of Exposure From External Sources in Restricted Areas," contains

requirements for control of access to high and very high radiation areas. It should

be noted that 10 CFR 20.1601(c) allows a licensee to apply to the Commission for

approval of alternative methods for controlling access to high radiation areas if the

licensee finds that the stated methods of control in the regulations would interfere

with utilization programs. The application should contain a description of the

proposed method along with a discussion of how the entrance or access point to

high radiation areas will be controlled.

Equipment and materials (e.g., anti-contamination clothing and respiratory

protection equipment) to protect personnel employed in the facility should be

discussed. The applicant should describe the facility conditions for which this

personnel protective equipment should be employed, and should also discuss

whether respirators should be used at the facility. The use of respiratory

protection equipment requires implementing and maintaining a respiratory

protection program in accordance with the requirements of 10 CFR 20, Subpart IL

If a respiratory protection program will be maintained, that program should be

described as it relates to the minimum program requirements of 10 CFR 20.1703.

The bases and values for the expected annual radiation exposure for all locations of

the facility should be discussed, including the exposure estimates for applicantemployed personnel, non-applicant-employed research and service personnel, and

visitors. This discussion should include the exposure limits and controls for such

groups as embryos, fetuses, declared pregnant women, minors, and students. The

plans and procedures for exposure control and dosimetry during the fill range of

normal facility operations, potential accident conditions, rescue and recovery, and

planned special personnel exposures (non-emergency) should also be discussed.

The applicant should describe the dosimetry used for assessing external radiation

exposures (e.g., whole body, extremities), including the frequency of dosimeter

readings, administrative dose action levels, and the suitability of the dosimetry

chosen with respect to the radiation sources anticipated and observed. The same

">

NLUREG- 1537, PART I 1 1-10 REV 0, 2/96
RADiATIoN PRaOEcnON PROGRAM AND WASTE MANAGEMENT

factors for how internal exposures and doses are assessed, evaluated, and

controlled should be described.

The applicant should describe the type of records retained to document the

conditions under which individuals were exposed to radiation. The applicant

should discuss the historical and current exposures to personnel and the associated

trends.

11.1.6 Contamination Control

The applicant should discuss the plans and bases of procedures for identifying and

controlling radioactive contamination, including methods established to assess the

effectiveness of the contamination control program. jThe discussion should include

information on the following topics, showing their relationship to regulatory

requirements and ALARA concepts:

* .a program for routine monitoring to detect and identify fixed and loose

contamination

* programs to control access to contaminated areas, avoid further spread of

contamination, and remedy contaminated areas

* personal monitoring and assessment of internal and external doses to

personnel occupying or entering contaminated areas, and methods for

appropriate surveying and "frisling. upon exit

* use of anti-contamintion techniques to protect workers, and control and

disposition of possibly contaminated clothingand materials

* procedures for monitoring and handling contaminated equipment and

components outside of contaminated areas that have not been

decontaminated

* criteria for classifying contaminated material, equipment, and working

areas, and managing, controlling, storing, and disposing of identified

contanination

* training programs for staff and visitors on'the risks of contamination and on

techniques for avoiding, limiting, and controlling contamination

* recordkeeping for contamination events, both for personnel and for

locations, including records to be available for facility maintenance and for

eventual decommissioning

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* the bases'of technical specifications, if needed, applicable to contamination

control: for exaonple, limits on storage and handling of radioactive sources,

especially unsealed ones; limitations on encapsulation of irradiated

materials; and use of fume hoods and hot-waste drains

11.1.7 Environmental Monitoring

The applicant should describe the environmental monitoring program, including

information relating to the following:

* verification of compliance 'with commitments made in environmental

reports, or other documents, if applicable; discussion of any standards used

in the environmental monitoring program

* for established programs, evaluation of the effectiveness of the program

* identification of potential facility impacts on the environment and the

evaluation of the need for remedial action or mitigation measures

* establishment of baselines for environmental quality, including data

comparing preconstruction or preoperational with operational

environmental monitoring results

The applicant should describe the written plans and the bases of procedures for

implementing the environmental monitoring program, and should discuss the

document control measures employed to ensure that the plans and procedures,

including changes, are reviewed for adequacy and approved by authorized

personnel, and are distributed to and used at the appropriate locations throughout

the facility.

The environmental surveillance program and its bases should be described. Air,

water, and land environments should be specifically discussed. These discussions

should include information on at least the following topics:

* probable facility-related contaminants and pathways to people

* selection of sampling materials and locations

* sample collection methods and frequency of collection

* sample analyses (analytical techniques) and sensitivities (detection limits)

* records of results and trends

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NUREG-1537, PART I X11-12 REV 0, 2196
RDIATION PROTECrION PROGRAM AND WASTE MANAGEMENr

11.2 Radioactive Waste Management

Each facility that is licensed to operate or utilize'a non-power reactor should

establish a program and procedures that are designed to ensure that radioactive

waste materials are identified, assessed, controlled, and disposed of in

conformance with all applicable regulations and in a manner to'protect the health

and safety.of the public and the environment. The magnitude and nature of the

effort required should depend upon the size and complexity of both the reactor

facility and its utilization programs. -Therefore, the nature and details of the

radioactive waste management program should also be commensurate with those

factors.- As noted previously, management of radioactive wastes could be an

auxiliary function assigned to existing personnel, such as people engaged in

radiation protection or operations. Earlier sections'of this chapter have addressed

the program and procedures for controlling and assessing radiation exposures and

doses at the facility due to all radiation and radioactive sources. In this section, the

applicant should address the program and procedures for further managing sources

classified as radioactive waste.

11.2.1 Radioactive Waste Management Program -

In this section, the applicant should discuss the philosophy and objectives of the

program for managing radioactive waste. The applicant should describe the

organizational structure within which it will administer the reactor-related

radioactive waste management program, including the organization and staffing

levels, authorities and responsibilitiesiand position qualifications. The worldng

relationships between such facility organizations as radiation protection and

operations stafl, and the standards, charters, procedures, or other documents that

specify the authority, duties, and responsibilities of the personnel in the radioactive

waste management organization should be discussed.' The policy governing the'

program, the allocation of policymaking responsibilities, and the administrative

plans and procedures that implement the facility policy should be described. The

overall program and how the organization,' policy, and program lead to effective'

management of radioactive waste should be evaluated and described.

The applicant should descnrbe the purpose, organization, and functions of any

committees assigned responsibility for overseeing radioactive'waste management.

The description should include each committee's charter, responsibilities, frequency

of meetings, audit and review responsibilities,- scope of any audits or reviews, and

qualifications and requirements for committee members.- How each committee's

work relates to the waste management organization and how they work together

should be discussed. If this information has already been described, reference that

discussion.

REV. 0. 2/96 -

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O. 2J96 -- -11-13 - -
CHA I I'.

The applicant should describe the waste management training program This

description should include the scope of fiLcility waste management training, as well

as specific training requirements for personnel associated with the operation and

use of the facility.

The applicant should describe the document control measures that ensure that the

plans and procedures involving radioactive waste, including changes, are reviewed

for applicability, approved by authorized personnel, and distributed to and used at

the locations where waste management activities are conducted.

The applicant should describe the scope of waste management reviews and audits.

This description should include the authority of waste management review apd

audit teams, the objectives and purposes of reviews and audits, and the bases for

scheduling these reviews and audits.

The applicant should describe the radioactive waste management recordkeeping

process, including retention periods, accessibility, review, and archiving, and

should discuss any special review of waste management records for accuracy and

validity. Records of radioactive wastes stored for the life of the facility or buried

on site should be discussed, as well as records for trend analysis.

The bases for any technical specifications related to the radioactive waste

management program should be described.

11.2.2 Radioactive Waste Controls

The applicant should discuss the definition of radioactive waste, the point in any

process that a radioactive component or material becomes classified as waste, and

the criteria for defining such waste. The applicant should describe the waste

management program procedures which ensure that radioactive wastes are

identified and characterized appropriately, as noted above, and the bases of the

procedures which ensure that radioactive wastes are adequately segregated from

nonradioactive wastes. The plans and procedures for managing all forms of

radioactive wastes generated during operations, research, and utilization of the

reactor should be described. Since radioactive wastes are radiation sources, they

should be described, along with other such sources, in Section 11.1 of the SAR

The applicant should describe the plans and bases for procedures for managng

gaseous and other airborne radioactive wastes generated during operations,

research, and utilization of the reactor, and radioactive waste ofl-gas collection

systems designed to be utilized at the facility. The function and the location of

each off-gas collection system should be described. At many non-power reactors,

the system for removing gaseous radioactive waste is integral to the ventilation

system for the facility and may have engineered safety functions. If these systems

NIJREG-1537, PART 1 11-14 Rsv 0,2/96

NIUREG-1537, PART I 1 1-14 REV 0, 2196
RADIATiON PRoTEaON PROmtAm AND WASTE MANAaEMENT

have been described in other chapters of the SAR, reference may be made here to

those discussions. For all off-gas and ventilation systems, the applicant should

describe the wastes produced by operation of the systems. 'Such items as filters

and scrubbers, which collect and concentrate wastes, should be discussed to

indicate the disposition of the radioactive material upon regeneration orreplacement. If the radioactive materials enter other waste treatment systems, the

applicant should indicate how suchltiansfers are made and note any possible

chemical or radiological effects of the transfer. The operation of any gas-clening

equipment and its designed'performance should be discussed in this section. The

bases of any, applicible technical specifications that control these functions should

be given. Also, the applicant should describe all secondary radioactive residues

that are generated during process treatment, their chemical and physical

composition, and the modes for handling, controlling, and storing them.-

The applicant should describe how liquid radioactive wastes are generated and

where they enter the waste control and treatment systems. Such items as

laboratory wastes, liquid spills, and cleanup solutions, including detergent wastes,

should be discussed. Information about the projected inventory levels, interim and

-long-term storage, and processing oftho'se streams to achieve volume reduction or

solidification should be included. This discussion should include information about

coolant cleanup systems and resin regeneration solutions and wastes, if applicable.

The objectives of the processes designed to treat radioactive or mixed liquid

wastes should be described. Any backup and special safety features designed to

ensure that the radioactive waste is contained during treatment should be

described. The designed eq'uipment and systems should be described, and

appropriate engineering drawings to showthe location ofthe equipment, flow

paths, piping, valves;'instrumentation,"and other physical features, should be

included, along with information on all features, systems, or special handling'

techniques that prevent uncontrolled releases or personnel exposures.

The applicant should describe the plans and procedures for managing solid

radioactive wastes generated during operations, research, and utilization of the

reactor. This description should include how solid radioactive materials are'

generated and where they' enter the waste control and treatment systems. For solid

radioactive wastes retained or stored on site for the life of the facility, the applicant

should discuss the control methods used. Integrity and corrosion characteristics

and the monitoring of the containment should be discussed, as well as the plan for

disposing 'of these radioactive wastes when the facility is permanently

decommissioned.

The applicant should describe the systems and equipment selected for identifying,

segregating, and safely managing the solid, liquid, and gaseous radioactive waste

that is generated, and should include appropriate engineering drawings showing

R~v.0,296 1-15STANARDFORMT AD COTE.

REV.0, 2/96 II1-is STAmAm FmmAT Am ComENr
CHAPE1R .I

the location of the equipment and associated features used for volume reduction,

containment, and/or packaging, storage, and disposal The applicant should also

discuss the bases of procedures associated with operating treatment equipment,

including performance tests, process limits, and the means for monitoring and

controlling to meet these limits. The bases of applicable technical specifications

that control these' procedures and functions should be discussed. The methods and

agents planned for all activities involving routine disposal or release to the

environment of radioactive wastes generated in the facility should be described, as

should methods used for packaging and shipping solid and liquid radioactive

wastes to other facilities or other means for processing, storage, or other

disposition.

The applicant should describe the program for minimizing radioactive waste for the

facility with respect to the following topics: (1) the specific numerical goals for

reducing the volume or radioactivity of each waste stream; (2) the periodic

assessments of reactor operations and experimental or utilization activities to

identify opportunities to reduce or eliminate the generation of wastes; (3) the

continuing efforts to identify and, where cost effective, implement waste reduction

technologies; and (4) any periodic independent reviews performed to evaluate the

effectiveness of programs to nminimize radioactive waste.

11.2.3 Release of Radioactive Waste

The applicant should identify all radioactive waste materials for which controlled

release to the environment or transfer to other parties for disposal is planned. This

discussion should include the projected concentrations, forms, chemical

compositions, and annual quantities of radioactive waste released under normal

operating conditions.

All points from which radioactive waste effluents are designed to be released from

the fiacility to the environment should be identified, using a site map to locate the

effluent release points and effluent monitoring equipment. Discussions and

detailed analyses of potential radiological impact of radioactive waste effluents and

the bases for continuous or intermittent monitoring should be provided in the

earlier sections of Chapter 11. For liquid releases to the sanitary sewerage, the

applicant shall ensure that the requirements of 10 CFR 20.2003 are met. The

applicant should describe the systems and procedures designed to ensure that

doses resulting from releases of radioactive effluents do not exceed applicable

regulatory limits and ALARA goals.

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RADIAON PROTECnON PRoaRAM AND WAsTE MANAGEMENT

11.3 Bibliography

American National Standards Institute/American Nuclear Society, ANSVANS

15.7,- Research Reactor Site Evaluation," -ANS, LaGrange Park, Illinois, 1977.

American National Standards Institute/American Nuclear Society, ANSIIANS

15.11, IRadiation Protection at Research Reactor Facilities," ANS, LaGrange

Park, Illinois, 1993. -

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.19, "Shipment and Receipt of Special Nuclear Material by Research Reactor

Facilities," ANS, LaGrange Park, Illinois, 1991.

Code of Federal Regulations, Title 10, "Energy," and Title 49, 'Transportation,"

U.S. Government Printing Office, Washington, D.C., revised periodically.

U.S. Nuclear Regulatory Commission, NUREG-085 1, "Nomograms for

Evaluation of Doses From Finite Noble Gas Clouds," January 1983.

U.S. Nuclear Regulatory Commission, NUREG/CR-2260 'Technical Basis for RG

1.145," 1981.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.109, "Calculation of

Annual Doses to Man From Routine Releases of Reactor Effluents."

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.145, 'Atmospheric

Dispersion Models for Potential Accident Consequence Assessments at Nuclear

Power Plants.'

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.9, "Acceptable

Concepts, Models, Equations, and Assumptions for a Bioassay Program."

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.9, proposed

Revision 1, "Interpretation of Bioassay Measurements," Task DG-8009.

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.10, "Operating

Philosophy for Maintaining Operational Exposures As Low As Is Reasonably

Achievable."

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.13, Revision 2,

"Instructions Concerning Prenatal Radiation Exposure."

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.29, 'Instruction

Concerning Risks from Occupational Radiation Exposure."

RE_ 0_/9 TNADFRA _11 N OT

REV. 0,2/9 11-17 STANDARD FoRmAT AND CoNTENT
CHAoiR I

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.29, proposed

Revision 1, "Instruction Concerning Risks From Occupational Radiation

Exposure,' Task DG-8012.

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.37, "ALARA Levels

for Effluents From Materials Facilities."

' -I

NUREG-1537. PART I 11-18 REv. 0,2/96

NUJREG- 1537. PART I 11 -18 REV. 0, 296
12 CONDUCT OF OPERATIONS

In this chapter of the SAR, the applicant should describe and discuss the conduct

of operations at the reactor facility. The conduct of operations involves the

administrative aspects of facility operation, the facility emergency plan, the security

plan, the quality assurance plan, the reactor operator requalification plan, the

startup plan, and environmental reports. The administrative aspects of facility

operations are the facility organization, review and audit activities, organizational

aspects of radiation safety, facility procedures, required actions in case of license

or technical specification violations, reporting requirements, and recordkeeping.

This chapter of the SAR forms the basis of Section 6 of the technical

specifications.,

12.1 Organization

In this section of the SAR, the applicant should discuss the organizational

structure; responsibilities, and staffing, including selection and training of

personnel. This discussion should show that the management and staff of the

facility are knowledgeable about the technical requirements to operate a safe

failibty, are responsible for complying with regulation's and license conditions, and

will implement a meaningful radiation protection program that will protect the

health and safety of the public, the facilityus, and the staff. Additional

information on these topics is given in Chapter 14, 'Technical Specifications," of

this format and content guide.

Not all owners and operators of non-power reactors have the same management

organization or office titles. Regardless of the details of theimanagement

organization or the complexity of the facility, the administrative functions-that

should be in place at a non'-power reactor facility are 'consistent among the various

non-power reactor designs. '

12.1.1 Structure

The applicant should discuss the organizational structure for the facility and should

submit an organization chart. Lines of responsibility and lines of communication

between groups should be'shown in a multilevel chart. The individual or group

w~ithlegal responsibility for holding the reactor license (eg., the university provost

or the dean) should be shown at the top of the organization, and the individuals

who operate the reactor should be shown at the bottom. Intermediate levels

should show the individuals who are in charge of the reactor facility and'reactor

operations (e.g., the facility director and the operations manager). The description

of the organizational 'structure should include the radiation safety function' and

indicate how the staff inplementing'that function interacts with 'the staff -

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-, REV.0, 2196 -1 . 12-1 STAN DAm FoRmAT AND CoNnENr
CHAPTER 12

responsible for reactor operations and the top administrative officials. The

multilevel chart should show the relationship of the review and audit function to

the organizational structure. The persons implementing the review and audit

function should communicate with the management of the reactor facility but

should report to an organizational level above this management to ensure

independence of the review and audit function.

12.1.2 Responsibility

The applicant should discuss the individuals or groups that'appear in the

organizational structure and their respective responsibility for the safe operation of

the reactor and the reactor facility, the protection of the health and safety of the

public and the workers at the facility, and the protection of the environment.

12.1.3 Staffing

The applicant should discuss staffing issues and the minimum staffing of the facility

when the reactor is not secured. The applicant should also-discuss the availability

of senior reactor operators'during routine operations and should list the events that

require the presence of a senior reactor operator at the facility. Staffing shall meet,

at a minimum, the requirernents of 10 CFR 50.54(J), (), (k), (), and (m)(l). The

applicant should show that these requirements are met.,

12.1.4 Selection and Training of Personnel

The applicant should discuss the selection and training of personnel. If minimum

requirements exist for the facility staf they should be discussed in this section.

For example, the facility director may have to meet certain educational standards.

The applicant should discuss the training programs at the facility, including the

initial training and the requalification training of reactor operators. The applicant

and licensed operators shall comply with 10 CFR Part 55. The applicant shall meet

the requirements of 10 CFR Part 19. The applicant should discuss the training

necessary to ensure that all workers meet the requirements of 10 CFR Part 19.

Specialized training may be needed for researchers who work near neutron beams

or whose experiments may cause changes in reactivity in the reactor. American

National Standards Institute/American Nuclear Society (ANSVANS) 15.4-1988

contains additional guidance on the selection and training of personnel for research

reactors.

12.1.5 Radiation Safety

The applicant should describe the organizational aspects of the radiation safety

program. The program itself should be discussed in detail in Chapter 11,

'Radiation Protection Program and Waste Management," of the SAR. The

NURE-157, PRT 122 R~. 02/9

NUREG- 1 537, PART I 12-2 REV. 0, 2/96
CONDU= OF~ OPEL4TMONS

radiation protection siaff can be part of the reactor facility staff or may be provided

as a service by a university or a corporation-wide group. The radiation protection

staff can report to managers of either the university or the reactor. If this staff

reports to the reactor manager, the applicant should discuss the method of raising

concerns to the level of the review and audit committee or to managers above the

facility management. The applicant should specifically discuss the authority of the

radiation safety'staff to interdict orierminate safety-related activities.

Additional guidince for radiation safety programs at non-power reactors is given

inANSTIANS 15.1-1993.

12.2 Review and Audit Activities

In this section of the SAR, the applicant should discuss the review and audit

activities at the reactor facility. This independent oversight is very important to

the safe operation of the facility and the protection of the health and safety of the

public.' NRC expects review and audit programs to be viable and to be strongly

supported by the licensee. Having independent experts examine certain activities

improves the quality of the program. Independent audit allows the licensee to find

and correct problems before NRC discovers them. Under its enfoircem'ent

program, NRC looks favorably on licensees that aggressively seek out and correct

problem~s in their programs. '

The committee established for the review function may be 'assigned approval

authority by the facility manager, or the facility manager may retain'the authority.

The applicant should explicitly state who holds the approval authority and should

specify the committee's authority and how it communicates and interacts with

facility management and university or corporate management.

12.2.1 Composition and Qualificafions

The applicant should discuss the composition and qualifications of the review and

audit committee. Both functions can be performed by one committee or two

separate committees. The applicant should discuss the minimum number of

committee members required. Committee members can be members of the reactor

facility staff as long as they constitute a minority of the quorum for voting. The

applicant should also -discuss the qualifications of committee members'. ' Because it

is sometimes difficult, especially for small facilities, to'have all types of engineering

and health physics expertise on'staff, the'review and audit committee provides-'

access to expertise. ' The committee members should have a variety of

backgrounds to provide both additional expertise that the reactor staff miy not

have and to allow a second opinion in areas in 'which the reactor staff has

expertise.' The -applicant should discuss the use'of committee members from`

outside the university or corporation. It isdesirable to have some members on the f ; -. .

RLT. 0, 2.M , - _ . . 12-3- -.,. -1 .111, -STANDARD FoRmAT AND CoNrE2,,r
CHAPER 12

committee from outside the university or corporation to increase the independence

of the committee.

12.2.2 Charter and Rules

The applicant should discuss the committee's charter and rules, including the

number of times the committee meets, the way the committee conducts business,

the requirements for a quorum when voting, and the way the committee distributes

its reports and reviews to the applicant. It is important thatthe definition'of a

quorum ensures that the reactor staff does not form a majority of members present

for any vote.

12.2.3 Review Function

The applicant should list and discuss the items that must be' reviewed by the

committee. Additional information on the review function can be found in Chapter

14 of this format and content guide.

12.2.4 Audit Function

The applicant should list and discuss the items that must be audited by the

committee. In addition to audits by the facility committee, the licensee may

consider entering into a auditing agreement with other non-power reactor facilities

to bring in staff members from other non-power reactors to perform an audit. This

approach has been very productive at the facilities that have used it.' The applicant

should consider all aspects of facility operations for audit, including-the radiation

protection program (see Section 11.1.2 of this format and content guide) and the

laboratory program at the facility if the p'rogram is conducted under the reactor

license (see Section 9.5 of this format and content guide). The emergency plan,

the physical security plan, and the operator requalification plan should be specified

for audit, although the requirement for auditing these plans may be contained in the

plan itself.

12.3 Procedures,

In this section of the SAR, the applicant should discuss the use of procedures at

the facility. NRC does not usually review procedures as part of licensing reviews.

The applicant should discuss the basic topics that the procedures do or will cover.

If laboratory work is conducted under the reactor license, the applicant should

reference procedures applicable to this work. The applicant should discuss the

methodology used for developing procedures, including the approval process. The

applicant should also discuss the process required to make changes to procedures

including substantive and minor permanent changes, as defined in ANSI/ANS

15.1-1 990, and temporary deviations to deal with special or unusual circumstances

NUREG-1537. PART I 12-4 REV 0,2/96 NUREG- 1 537, PART I 1294 REV 0, 2196
- CONDUCT OF OPERAMONS

during operation. The applicant should note that 10 CFR 50.59'may apply to

changes to procedures. See Chapter 14 of this format and content guide for

additional information on the minimum acceptable set of procedures.

12.4 Required Actions

In this section of the SAR, the applicant should discuss actions to be taken in the

event of a violation of the facility safety limits or the occurrence of a reportable

event. In the case of a safety limit violation, the reactor shall be shut down, the

proper facility management notified,- the event investigated, and NRC notified.

Additional information on required actions for a safety limit violation is given in

Chapter 14 of this format and content guide..

The applicant should discuss what occurrences are considered reportable events

and the actions to be taken if a reportable event occurs. These actions usually

consist of returning the reactor to its normal condition or shutting the reactor.

down and reporting the event to facility management and NRC. Additional

information on reportable events is given in Chapter 14 of this format and content

guide.

12.5 Reports

In this section of the SAR, the applicant should discuss what information should be

reported to'NRC, the format of reports, the timing of reports, and the distribution

of reports to NRC. This discussion and related technical specifications on reports

take the place of the'reporting requirements in 10 CFR 50.72 and 73, which are

not applicable to non-power reactors. The applicant should discuss the annual

operating report and should list the required contents. The annual report may also

be used to meet the reporting requirements of 10 CFR 50.59 by discussing changes

made to the facility and procedures and the tests and experiments conducted under

the authority of 10 CFR 50.59. Additional information on the contents of annual

operating reports is given in Chapter 14 of this format and content guide.

The applicant should discuss special reports, which are used to inform NRC of the

violation of safety limits or the occurrence of reportable events, as previously

discussed.- The special reports are also used to inform NRC of changes in facility

management and of changes in the transient and accident analyses in the SAR.

Additional information on special reports is given in Chapter 14 of this format and

-content guide. - - .,

R~~v.~ TDR OMTADC]mr _,/61. 

REV. 0, 2196 12-5 STANDARD FoRmAT AND CONTENT
CHAPTER 12

12.6 Records

In this section of the SAR, the applicant should discuss facility records and

describe the recordkeeping system for the facility, the types of records that need to

be retained, and the period of retention. Additional information on records is

given in Chapter 14 of this format and content guide.

12.7 Emergency Planning

In this section of the SAR, the applicant should give a brief overview of the plan.

The reader should be referred to the emergency plan for additional details.

The applicant should follow the guidance of ANSI/ANS 15.16-1978, which is

endorsed and amplified by Regulatory Guide 2.6, March 1983, 'Emergency

Planning for Research and Test Reactors" (appears as Appendix 12.1). The

applicant should also review NUREG-0849, 'Standard Review Plan for the

Review and Evaluation of Emergency Plans for Research and Test Reactors'

(appears as Appendix 12.2).

The applicant should ensure that the terminology for releases discussed in the

emergency plan match that used in the revision of 10 CFR Part 20 that became

mandatory on January 1, 1994. Specifically, effluent concentrations (ECs) have

replaced maximum permissible concentrations (MPCs), although use of dose

values in millirem (millisievert) would be a more appropriate protective action

guideline because dose is the ultimate criterion specified. The applicant should

ensure that the action levels discussed in the emergency plan for each emergency

class follow the guidance. If it is impossible for an event at a particular facility to

reach a given action level, that emergency class is not possible and that fiact should

be so stated in the plan

12.8 Security Planning

In this section of the SAR, the applicant should briefly discuss security planning.

The information in the SAR must be nonproprietary and must not contain

safeguards information. The proprietary or safeguards version of the security plan

is protected from disclosure by the regulations. The guidance in Regulatory

Guide 5.59 (appears as Appendix 12.3) should be referred to when developing a

security plan If an applicant believes that the special nuclear material the facility

possesses or will possess will not allow the applicant to fall under the regulations

for material of moderate or low strategic significance, the applicant should dontact

the project manager.

NUREG-1537, PART I 12-6 REV. 0, 2./96

NUREG-1537,PsAT I 12-6 REV. 0, 2/96
CoNn~ucr oi OPERATioNs

12.9 Quality Assurance

Section'50.34(a)(7) of 10 dFR requires that applicants for construction permits to

build non-power reactors describe a quality assurance program for the design and

construction of the structures, systems, and components of the facility. Section

50.34(bX)(6ii) requires a description in the SAR of managerial and administrative

controls to be used to ensure safe operation: The applicant should consider the

guidance in Regulatory Guide 2.5 (appears as Appendix 12.4) and ANSI/ANS

15.8-1976 in developing quality assurance programs for non-power reactors.

These guidance documents provide an acceptable method of complying with the

program requirements of 10 CFR 50.34.

12.10 Operator Training and Requalification

Bach reactor operator or senior reactor operator is required to successfmlly

complete a requalification program developed by the licensee that has been

approved by the Commission by passing a comprehensive written examination and

an annual operating test. The operator requalification plan should describe the

operator requalification program to be administered to non-power reactor

personnel posessing operator and semor operator licenses in order forthemto

maintain active status. The plan should also discuss the requirements for reactor

operators and senior reactor operators to maintain active status arid the steps that

should be taken to return an inactive operator to active status. The operator

requalification plan is normally aiseparate docurment. It should include the

organizational structure for implementing the program.

Section 50.54(-1) of 1O CFR requires itht within 3 months after an operating

license is issued, the licensee have in effect an operator requalification program,

which at a minimum meets the requirements of 10 CFR 55.59(c). In accordance

with 10 CFR 55.59(a)(1) and 55.59(c)(1), the requalification period should not

exceed 24 months.

The operator requalifcation plan should include information on the following:

:* requalification schedule'' '

* lectures, reviews, and examinations-

* on-the-job taining

* emergencyprocedures - :

* inactive operators

.: :* -evaluation and retraining of operators

* requalification documentation and records

requalification document review and audit

REv. 0,2196 12.7 STAIDA]W FORMAT AD CONTENT

RLN'.O.2/96 12-7 STAmARD FoRmAT Am ComaqT
CHAPTER 12

Preplanned lectures should be given on a regular and continuing basis as required

by 10 CFR 55.59(c)(2). Some examples of lecture topics are the following:

* nuclear theory and principles of operation

* facility design and general and specific operating characteristics

* instrumentation and control systems

* reactor protection system

* engineered safety features

* normal, abnormal, and emergency operating procedures

* radiation control and safety

* technical specifications and applicable portions of 10 CFR

* other facility specific training

Self-study methods may be considered an adequate and appropriate training

method for the lecture program topics when learning objectives are properly

measured by examination or documentation of expertise. Self-study programs

must be justified by the applicant.

On-the-job training should occur during the requalification period so that each

operator (1) is involved in facility manipulations [see 10 CFR 55.59(c)(3)(1)],

(2) understands the operation of apparatus and mechanisms associated with

control manipulations and knows operating procedures [see 10 CFR

55.59(c)(3)(Ji)], (3) is cognizant of changes in ficility design, procedures, and

license Esee 10 CFR 55.59(c)(3)(iii)], and (4) reviews the contents of all abnormal

and emergency procedures on a regular basis [see IOCFR 55.59(cX3)(Qv)].

Reactor operators must perform and senior reactor operators must either perform

or supervise the appropriate items from 10 CFR 55.59(cX3)(1), such as reactor

startup, shutdown, and significant power change, on an annual basis.

NRC has approved requalification plans for on-the-job training that are similar to

the following:

* Over the 2-year requalification period, each licensed individual should

perform at least ten reactivity control manipulations in any combination of

reactor startups, shutdowns, or significant reactivity changes, and each

licensed operator should perform at least one reactor startup and one daily

checkout quarterly at intervals not to exceed 4 months.

* Any changes in procedures, technical specifications, regulations, as well as

any change with safety significance to the facility should be reviewed by

every licensed operator on an ongoing basis as part of a required reading list.

* Each operator should participate in at least half of the emergency drills per

year specified in the technical specifications. Operators who did not meet

NUREG-1S37, PART 1 12-8 REv. 0,2/96

NUREG-1537,PAPT 1 12-8 REV. O. 2/96
CONDUCr OF OPATIONS

the minimum drill requirement should receive special training on proper

response to emergencies and a documented review of the last drill missed as

well as a walkthrough of the facility related to proper emergency responses.

The drill and applicable emergency procedures should be reviewed with all

licensed operators within 30 days after completion of the drill.

The requalification program should include provisions for evaluating operators

[see 10 CFR 55.59(c)(4)] that include written examinations, observation and---

evaluation of operator performance, and simulation of emergency or abnormal

conditions. The applicant should discuss provisions for accelerated requalification

if performance evaluations indicate the need [see 10 CFR 55.59(c)(4)(v)].

Examinations should be administered as necessary to assess the progress of the

lectures or self study. A comprehensive requalification written examination is

required for all operators at least biennially [see 10 CFR 55.41, 55.43,

55.59(a)(2)(I), and 55.59(c)(4)]. The contents of the comprehensive examination

should be listed and should cover all lecture topics and requirements of the -

regulations.

For example, NRC has accepted requalification plans similar to the following:

* The acceptance criterion for all graded examinations is 80 percent and all

operators are required to complete each examination satisfactorily. -

* A score on the written or other examination equal to or greater than,.

80 percent may require no additional training. Nevertheless, the results of

all examinations including missed questions should be reviewed with the

operator to ensure proper understanding.

* A score on the written or other examination in the range of 65 to 79 percent

requires additional training on those areas or topics where weaknesses or

deficiencies are indicated. This retraining and retesting are completed within

60 days from the date the examination was administered and before the

candidate is considered requalified. In this case the candidate need not be.

removed from licensed duties subject to the evaluation of the reactor

manager ora duly authorized representative.. -. - -

A score on the written or other examination of less than 65 percent requires

that an evaluation be performed by the facility director or designated

representative within I month. The evaluation is to determine if the

deficiencies require that the individual be removed from licensed duties

pending completion of any accelerated retraining. In any case the licensed

operator is removed from licensed duties if within 4 months he or she does

not achieve a passing grade after reexamination..

REv. O, 2/96 12-9-- STANDARD FORMAT AND CoIEn
CHAPSt 12

* Regardless of the score, if the individual's test indicates a deficiency in a

critical area'that affects safety, training is promptly administered to correct

the deficiency or the'operator is removed from licensed duties in the affected

area until the deficiency is corrected.

Each reactor operator and senior reactor operator is required to take an annual

operations test to demonstrate operational proficiency and understanding of system

responses [see 10 CFR 55.45(a)(2-13), 55.59(a)(2Xii), and 55.59(c)(4)]. NRC has

approved requalification plans similar to the following:

e Each licensed reactor operator and senior reactor operator demonstrates

satisfactory understanding of the operation of the facility systems, operating

procedures, and license as well as changes in facility procedures and the

license during an annual walkthrough examination administered by a

designated senior reactor operator.

* The annual operations test and the annual wallthrough examination are key

factors in evaluating the continued competence of the licensed operator for

demonstrating both (1) operational proficiency and understanding of system

responses and (2) overall satisfactory understanding of the operations of the

facility, operating procedures, and facility license changes. The results of

these two examinations are used as primary input for evaluating operator

performance for requalification purposes.

* An indepth evaluation of the operating performance of each licensed

operator is performed and documented biennially, at a minimium, by a

summary and judgment statements. The biennial evaluation includes results

from the written examinations, the annual' operations test, the annual

wallkthrough examination, and other on-the-job evaluation of operational

proficiency as well as any other available indications of the operator's

capability to discharge duties in a safe and competent manner, including

participation in practical and special training, instructional activities, and

other work activities.

The requalification plan should discuss records associated with the program [see

10 CFR 55.59(c)(5)]. NRC has accepted requalification plans similar to the

following:

* Operator requalification records are kept to ensure that all requirements of

the facility plan'are met. Each operator has an individual folder or notebook

containing signature blocks for lectures attended, prepared or'assigned selfstudy sessions, reactivity manipulations performed, weekly and daily

checkouts performed,' and quarterly emergency drills participated in. The

notebook also contains copies of written examinations administered, the

NUREG-1537, PART 1 12-10 REV 0,2/96

NUREG- 1537, PART I 12-10 REV 0, 2/96
CCoNDucr oF OPERAnoNs

answers given by the operator, results of any evaluations, and

documentation of any additional training administered in areas in which an

operator has exhibited deficiencies. 'The performance of, or participation in,

special training such as for fuel handling or use of emergency equipment is

also logged in the notebook.

A master requalification training manual is used to organize training

requirements; the manual contains a schedule of all required lectures',

reviews, emergency drills, and other exercises. The date the item is

performed is indicated in this manual. A section of this manual is designated

to contain completed training items, attendance sheets, master copies of

tests given, and lecture outlines if available. - ' .

*- Required documents and records pertaining to the requalification program

are maintained as part of the facility records for at least 6 years. The records

, including the master training file are retained for each reactor operator or

senior reactor operator until the respective operator's license is renewed or

surrendered.

To maintain active status, 10 CFR 55.53(e) requires that each licensed reactor

operator or senior reactor operator actively perform the functions of a reactor

operator or senior reactor operator for a minimum of 4 hours each calendar:

quarter. For senior reactor operators, direct supervision of these operations may

be considered equivalent to actual performance. If this requirement is not met, the

license becomes inactive; before reactivation of the license, the licensee should

verify that the qualifications and status of the operator are current and the operator

should perform 6 hours of licensed activities in the position that is being recertified

under the direction of a licensed operator or senior reactor operator as

appropriate. For example, NRC has approved requalification plans sixilar to the

following:

An operator who has not been actively performing licensed functions for a

period in excess of 4 months is required to denonstrate to the reactor

manager or duly authorized representative that his or her knowledge and

understanding of the operation and administration of the facility are:

satisfactory before returning to licensed duties. An interview and evaluation

or a written, oral, or operational examination, or a suitable combination, can

be used.

* Any deficiencies uncovered are corrected before the individual resumes

performance of licensed functions.

* The operator performs 6 hours of operation under the direction of a reactor

operator or senior reactor operator as appropriate.

REv. 0,2/96 12-Il STANDARD FORMAT #.n CoNTr

RE Ix%. 0, 2/96 12-1 1 STANDmD FORMAT Aim CONTENT
CHAPTER 12

The applicant should discuss audits of the requalification plan in the plan or as part

of Section 6 of the technical specifications. The plan should be audited by the

facility review and audit committee at least every other calendar year, with the

interval between audits not to exceed 30 months. The audit should consist of a

sampling of the plan records for completeness and compliance with the

requirements of the plan.

12.11 Startup Plan

In this section of the SAR, the applicant should describe the startup plan for the

faciity. This section is applicable to new facilities or license amendments

authorizing modifications that require verification of operability before normal

operations are resumed. Startup plans ensure that the operating characteristics are

well understood and validate the predicted behavior of the reactor. Measurements

of selected parameters of the reactor should be compared to calculated values to

verify analytical methods and ensure that meaningful acceptance crita for the

reactor have been established from the calculational methods. The acceptance

criteria should ensure that the reactor is functioning within the bounds for which it

was designed and analyzed and that the license and the technical specifications are

satisfied.

The plan should include the following:

* a well-planned systematic set of subcritical multiplication measurements or

an inverse multiplication approach to critical measurement during fuel

loading, and confirmation that subcritical multiplication or critical fuel

loading is within preestablished acceptable limits

* an experimental measurement plan to determine the important operational

reactor physics parameters (such as control rod worth, excess reactivity,

reactor thermal power, coefficients of reactivity, and power pealdng factors)

and thermal-hydraulic parameters (such as fuel, cladding, and coolant

temperatures, reactor coolant system flow rates, and pressure drops, if

appropriate); comparis'ons with predictions and acceptance criteria; and

investigation and discussions of arny discrepancies that may have arisen

* measurements of magnitudes of area radiation fields and radioactive

effluents, and comparisons with predictions in the SAR and preestablished

acceptance criteria

* measurements of performance of other systems (e.g., ventilation, engineered

safety features, or emergency power), and comparisons with predictions in

the SAR and preestablished acceptance criteria

NUREG-1537, PART I 12-12 REV. 0,2/96

NUREG-1537,PsATI1 12-12 REV. 0, 2/96
CONDUCT OF OPERPAIONS

The applicant should submit a startup report to NRC in accordance with the

timeframe required in the technical specifications. A sample outline for a startup

report follows (the actual contents of the report will vary by reactor type and

scope of the report):- :

measurements and comparisons with the prediction of subcritical

multiplication for initial core fuel loading

* .critical mass and final criticality conditions for the initial core and

operational core, including comparisons with acceptance criteria

preestablished in the SAR calculations and analyses

* control and regulating rod calibration, including measurements of differential

and integral rod worth for the initial and operational core, and comparisons

with acceptance criteria preestablished in the SAR calculations and analyses

@ excess (operational) reactivity, including comparisons with acceptance

criteria preestablished in the SAR calculations and analyses

* measured shutdown margin, including comparisons with acceptance criteria

preestablished in the SAR calculations and analyses:

* reactor power calibration, including methods and measurements that ensure

operation within the license limit; discussion of nuclear instrumentation

setpolnts, detector positions, and detector output; and comparisons with

acceptance criteria preestablished in the SAR calculations and analyses

* thermal neutron flux distributions, including comparisons with acceptance

criteria preestablished in the SAR calcltions and analyses

* radiation measurements of reactor coolant fission product inventory or:-

release during the startup test program to detect potential degradation of the

.fiel fission product barrier or contamination from other sources; results of

radiation measurements to show effectiveness offacility shielding;

measurements of airborne effluents released from the facility; and

comnpansons with acceptance criteria preestablished in the SAR calculations

and analyses

* reactivit wofrbs of experimen facilities, including comparisons with

acceptance criteria preestablished in the SAR calculations and analyses

* results of determination of temperature coefficients and void coefficients for

the reactor core, and comparisons with acceptance criteria preestablished in

the SAR calculations and analyses

REV. 0,2/96 12.13 STANDARD FouMAT CoImT

REV. 0,2196 * 12-13 STANDARD FORMAT AND CoNTENT
CHArS 12

* Thermal-hydraulic characteristics such as fuel, cladding, and coolant

temperatures, and reactor coolant system flow rates and pressure drops

* measurements of performance of engineered safety features and other tested

systems, including comparisons with acceptance criteria preestablished in the

SAR calculations and analyses

This list is not complete. The components and systems to be tested and the test

results reported to NRC will vary with reactor design and the subject of the

licensing action.

12.12 Environmental Reports

* NRC must comply with the National Environmental Protection Act of 1969, as

amended, and recognizes a continuing obligation to conduct its domestic licensing

in a manner that is receptive to environmental concerns. This means that licensing

actions for non-power reactors must address environmental concerns. These

licensing actions usually fall into one of three categories: those identified as

categorical exclusions, those requiring the preparation of an environmental

assessment, and those requiring the preparation of an enviromriental impact

statement. Generally, most license amendments for non-power reactors fall under

one of the categorical exclusions identified in 10 CFR 51.22; however, this is not

always the case. Issuance of construction permits or operating licenses and license

renewals for test reactors falls under 10 CFR 51.20 as regulatory action requiring

environmental impact statements. Construction permits, operating licenses, license

renewals, decommissioning plan orders, and license termination orders for research

reactors are usually actions requiring environmental assessments. The

environmental assessment is used to determine if an environmental impact

statement or if a finding of no significant impact should be prepared.

For those actions for which an environmental impact statement or an

environmental assessment must be prepared by the staff NRC may require the

applicant to submit an environmental report to help the staff to comply with the

environmental regulations. The requirements for the contents of an environmental

report are given in 10 CFR 51.45. For example, for an initial application for a

construction permit and operating license or for license renewal, the applicant

should describe the facility, discuss the environmental effect of facility site

preparation and construction, environmental effects of facility operation,

alternatives to construction and operation of the facility, and costs and benefits of

facility alternatives.

K-i

NIJRBG-1537. PART I 12.14 REV. 0,2/96 NUREG-1537,PARTI 12-14 REV. O. 2/96
CoNDucT oF OPERAnoNs

12.13 References

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.1, 'The Development of Technical Specifications for Research Reactors," ANS,

LaGrange Park, Illinois, 1990.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.4, 'Selection and Training of Personnel for Research Reactors," ANS,

LaGrange Park, Illinois, 1988.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.8, 'Quality Assurance Program Requirements for Research Reactors," ANS,

LaGrange Park, Illinois, 1976.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.11, "Radiological Protection of Research Reactor Facilities, ANS, LaGrange

Park, Illinois, 1993.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.16, "Emergency Planning for Research Reactors," ANS, LaGrange Park,

Illinois, 1978.

REV. 0,2196 12-15 STANDARD FOT AND CoNmwr

REV. 0, 2/96 12-15 STANDARD FORMAT AND CONTENT
Appendix 12.1

Regulatory Guide 2.6

Emergency Planning for Research

and Test Reactors

STANDARD FORMAT AiD CONTENT
Revision 1

U.S. NUCLEAR REGULATORY COMMISSION March 1983

vREG ULATORY .- GUIDE , -OFFICEOF NUCLEAR REGULATORY RESEARCH- i

REGULATORY GUIDE 2.6

(Task HF 2014)

EMERGENCY PLANNING FOR RESEARCH AND TEST REACTORS

A. INTRODUCTION

Paragraph S0.34(bX6Xv) of 10 CFR Part S0, Domesti

Lacensung of Production and Utilization Facilities," require

that - each apphcation for a license to operate a facilit

include in a Final Safety Analysis Report (FSAR), alon

with other informatLon, the applicant's plans for copin

with emergencies,: including the items. specified L

Appendix E, 'Emergency Planning and Preparedness fo

Production and Utilizaton Facilities," to 10 CFR Part 5C

Paragraph 50 54(q) requires hcensees -to follow an,

maintain in effect emergency plans that meet the require

ments of Appendix E to 10 CFR Part 50

wiThis regulatory gwude provides licensees and applicant

with a method acceptable to the NRC staff for complyin

with the Commission's regulations with regard to th

content of emergency plans for research and test reactors

Any guidance in this document related to unforms

tion coUection activities has been cleared under OMI

ClearanceNo. 3150-0011

B. DISCUSSION

Working Group ANS-1516 of the American Nuclea

Society Subcommittee ANS-15 has developed ANSI

ANS 15.16-1982, 'Emergency Planning for Researcl

Reactors," which is generally consistent with curren

regulatory requirements. Ths standard was developed ti

provide specific acceptance criteria for complying with th

applicable requirements set forth in § S0 S4 and in Appen

dix E to 10 CFR Part S0 These criteria provide a basns fo

research and test reactor licensees and applicants to develol

acceptable ndiological emergency response plans an

improve emergency preparedness at their facilities

The Commission's interest in emergency planning i

focused primarily on situations that may cause or ma:

*Cophes may be obtasned from the American Nucear Soaecty 555 North Kenuagton AMene. La Grlae Park. 111 e5S25.

threaten to cause radiological hazards affecting the health

and safety of 'the' public. 'Emergency plans should be

I- direted toward mitigang the consequences of emergencies

s and should provIde reasonable aissurance that appropriate

y measures can and will be taken to protect the health and

g -'safety of the public in the event of an emergency. Although

g i i it is not practicable to develop a completely detailed plan

n encompassing every -conceivable 'type of emergency

r situation, advance planning and provasions for ensuring the

* availability of necessary equipment, 'supplies, and servces

Id can create a high order of preparedness and ensure an

orderly and timely decisionmaking process at the time of an

emergency The plans should be an expression of the overall

concept of operation that describes how the elements of

s advance planning have been considered and the provisions

g that have'been made to cope'wnth emergency'situations

s In the judgment of the NRC staff, the potential radiological hazards to the public associated with the operauon

of research and test reactors arc considerably less than

B - those involved with nuclear power plants 'In addition,

because there are many different kinds of research and test

reactors, the potential for emergency situations arising and

the consequences thereof vary from facility to facility.

These differences and variations are expected to be reflected

r realistically an the emergency plans and procedures developed

1 for each research and test reactor facility.

h

t C. REGULATORY POSITION

e TherequirementsinANSI/ANS 15.16-1982. Emergency

I- Planning for Research Reactors," are generally acceptable

r to the NRC staff as a means for complying with the requirep ments in § 50 54 and in Appendix E, "Emergency Planning

d and Preparedness for Production ,nd Utilization Facilities."

to 10 CFR Part 50 as related to research and test reactors,

subject to the following clarifications.

y 1. Responsibility for planning and implementing all

emergency measures within the site boundaries rests with

the licensee. In this context, the site boundaries should be

USNRC REGULATORY GUIDES

Rogulatory Guides are issued to describe and rxke available to the public mrthots acceptable to the NRC Staff GI bnoiementlg apeciic Partt of the COmmISSlon1S r guitIoIs, to delineate tech. niQues used by the staff In evaluating specific problems or postu- lated acCidnts Of to Provide guidance to applicants Regulatory Guides are nof SubStitUte$ for reguiatlons. and compliance with tenm Is not required. Methods and solutions different from those set out In the guides will be acceptable If they Provile a basis for the findings reousite to the Issuance or continuance of a Permit or license by the Commission

This guide was Issued after consideration of comments received from the public. Comments and suggestions for improvements in these guioes are encouraged a: all times, and guides wili De revied. as aoorooriate. to accommodate comments and to reflect new inlorna- tlion or experience

Comments Should be sent to the Secretary of the CommissIon.

U.S. Nuclear Regutatory Conminsson. Washington. D.C. 20555. AttentIon Docx ting and Service Branch.

'he guides afe Issued In the following ten broad divisions,

I Power Reactors 6. Products 2. Research and Test Reactors 7. Transportation 3. Fuels and Materials Facliuies S. occupational lealth 4. Environmental and Siting 9. Antitrust and FInancial Review S. Materials and Plant Protection 10. General

CopleS of Issued guides may be purchased at the current Government

Printing Oftice price. A Subscription service for future guides In $te- cilic divisions Is available through the Government Printing Office information on the Wubscription service and current GPO Prices may be obtained by writing the US Nuclear Regulatory Commission.

Washington. D C 20555. Attention Publications Sales Manager
clearly defined. Supporting' organizations that would

augment the licensee's emergency organization, eg., fire

department, hospitals, and security org°nizations, should be

specified. Planning and timplementation of measures to cope

with reactorrelated emergencies beyond the site boundary

should be commensurate with and based on the potential

consequences of credible accidents or incidents' 'Te

emergency plan should describe this planning basis and the

corresponding arrangements and apeements among the

licensee and the local, State, or Federal agencies expected

to respond.

2. The radiation dose levels of the energency action

levels established for the various emergency dcaes are

slightly different from those specified for power reactors.

However, in the judgment of the NRC staff, the radiation

dose levels specified in Table I of the standard are adequate

for the credible accidents associated with the operation of

research and test reactors, and the specified action levels

provide reasonable assurance that protectve measures

associated with the action levels specified mn Table I can and

will be taken, provided appropriate emphasis is also given to

developing emergency action levels that relate directly to

facility parameters (eg., pool water levels and area

radiation monitors)

3. Emergency action levels related to facility pammeters,

effluent release levels, and equipment conditions should be

developed to the extent feasible for each emergency class

4. Details that can reasonably be expected to change

from time to time, eg., names and telephone numbers,

specific items of equipment and supplies, inventory ists,

and step-by-step procedures or checklists that may be

altered as a result of experience or test exercises, should not

be incorporated into the plans but should be listed in the

emergency implementing procedures.

S. Emergency procedures that implement the emergency

plan need not be incorporated into the plan but should be

listed by title in an annex to the emergency plan The

emergency implementing procedures should be maintained

and available at the facility for inspection and review at any

time by a representative of the NRC.

6. The procedural system used by the licensee for the

revnew and approval of emergency implementing procedures

should contain instructions governing the iting, revsSg,

and updating of nmplementing procedures. The instructions

should specify the- methods to be used to ensure that

procedures, revsons, and changes are reviewed for adequacy,

approved for use, and distributed to user organizations and

individuals having the responsibility for implementing the

procedures.-

D. IMPLEMENTATION

Except in those cases mi which an applicant or licensee

proposes acceptable alternative practices or methods for

complying with specified portions of the Commission's

regulations, the practices or methods described in this guide

wili be used as a basis for evaluating the adequacy of the

emergency plans and preparedness of applicants for a

license to operate a research or test reactor as well as the

plans and preparedness of current licensees for such reactors.

2 6-2

K-I
VALUE/IMPACT STATEMENT

1. THE PROPOSED ACTION

The licensee of a research and test reactor is required by

the Commission's regulations to develop plans for coping

with emergencies Specific guidance is needed to provide

acceptance criteria for complying with the applicable requirements set forth in § S0.54 and in Appendix E of 10 CFR

Part SO Regulatory Guide 2.6. "Emergency Planning for

Research and Test Reactors," provides basic guidance for

complying with the regulations More definitive guidance,

however, has been developed by the American Nuclear

Society Subcommittee ANS-IS in ANSI/ANS 15 16-1982.

"Emergency Planning for Research Reactors" The proposed

action would endorse this standard with appropnate aupplementary material in a revision to Regulatory Guide 2 6

1 1 Value/impact Assessment

The proposed action would provide licensees and applicants definitive guidance for developing emergency plans

that meet the appropriate regulation

Value - The .alue of the proposed iction sould be more effective emergency preparcdness around rcseArch and test

reactors Endorsmg a nutional consensus standard reduces the

expenditure of staff resources in developing the guidance

Impact - Most of the impact on industry has already

occurred during development. review, and approval of the

consensus standard and in attempting to Comply with the

upgraded emergency preparedness requirements promulgated in August of 1980 For those members of the

research reactor community that have not previously

upgraded their emergency plans, it is estimated that it will

take approximately 2 man-months to do so

1.2 Decision on the Action

Regulatory Guide 2 6 should be revised to endorse

ANSIJANS IS 16-1982

2. TECHNICAL ALTERNATIVES

Because the regulatory guide would endorse a consensus

standard, no technical alternatives have been considered.

3. PROCEDURAL ALTERNATIVES

Because ANSI/ANS 15.16-1982 is generally consistent

with current regulatory requirements, revising Regulatory

Guide 2.6 to endorse that standard was selected as the

appropriate procedural alternative

4. STATUTORY CONSIDERATIONS

4 1 NRC Authority

Authority for this action is derived from the Atomic

Energy Act of 1954. as amended, through the Commission's

regulations in Title 10, Chapter 1, of the Code of Federal

Regulations

4.2 Need for NEPA Assessment

Since the guidance in the proposed regulatory guide

revision does not represent a major action as defined by

paragraph 5l S(aXlO) of 10 (FR Part SI implementtion of

the regulatory guide does not require a NEPA assessment

S. RELAlONSHIPTO OTHER EXISTING OR PROPOSED

REGULATIONS OR POLICY

This revision to Regulatory Guide 2.6 relates to the NRC

emergency prepardness regulations, Regulatory Guide 1.101,

and NUREG-06S41FEMA-REP-1.

6. SUMMARY AND CONCLUSION

A revision to Regulatory Guide 2.6 should be published.

2 6-3
Appendix 12.2

NUREG-0849

Standard Review Plan for the

Review and Evaluation of Emergency Plans

for Research and Test Reactors

STANDARD FORMAT AND CONTENT
NUREG-0849

Standard Review Plan for

the Review and Evaluation of

Emegency Plans for Research

and Test Reactors

Manuscript Completed September 1983

Date Published: October I9S3

Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555

, A.r
ABSTRACT

This document provides a Standard Review Plan to assure that complete

and uniform reviews are made of research and test reactor radiological

emergency plans.

The report is organized under ten planning standards which correspond

to the guidance criteria in American National Standard ANSI/ANS 15.16 - 1982

as endorsed by Revision 1 to Regulatory Guide 2.6. The applicability of the

items under each planning standard is indicated by subdivisions of the

steady-state thermal power levels at which the reactors are licensed to

operate.

Standard emergency classes and example action levels for research and

test reactors which should initiate these classes are given in an Appendix.

iii
CONTENTS

Page

ABSTRACT............................................................ iii

ACKNOWLEDGEMENTS.............;...................................... vii

INTRODUCTION ........................................................ 1

CONTENT OF EMERGENCY PLAN ........................................... 3

AREAS OF REVIEW, PLANNING STANDARDS, AND EVALUATION ITEMS ........... 5

1.0 INTRODUCTION ................................................... 5

2.0. DEFINITIONS .................................................... 6

3.0 ORGANIZATION AND RESPONSIBILITIES ............................... 7

4.0 EMERGENCY CLASSIFICATION SYSTEM ................................ 10

5.0 EMERGENCY ACTION LEVELS ........................................ 12

6.0 EMERGENCY PLANNING ZONES ....................................... 13

7.0 EMERGENCY RESPONSE ............................................. .14

8.0 EMERGENCY FACILITIES AND EQUIPMENT ............................. 17

9.0 RECOVERY ....................................................... 19

10.0 MAINTAINING EMERGENCY PREPAREDNESS ............................. 20

APPENDIX I: EMERGENCY CLASSES, ACTION LEVELS,

AND PURPOSES ....................................................... 23

APPENDIX II: ALTERNATE METHOD FOR DETERMINING

THE SIZE OF AN EMERGENCY PLANNING ZONE ............................. 25

v
STANDARD REVIEW PLAN FOR THE REVIEW AND EVALUATION

OF EMERGENCY PLANS FOR RESEARCH AND TEST REACTORS.

INTRODUCTION

Safety analyses for research and:test reactors are-based on the concept

of .a postulated Design Basis Event (DBE), an event for which the risk to the

public health and safety is' greater"than'that from any'event that can.be

mechanistically postulated. The rationale for using the DBE'for.research

and test'reactors is to assess the potential effects to the public health

:and.safety and is based on the'determination that the offsite doses from the

DBE be within,:the 'requirements of 10 CFR Parts 20 and 100.' Consequently, if

the requirements are met fora.DBE condition,' then the capability of the .- facility to withstand normal and abnormal operational transients and a broad

spectrum of postulated credible' accidents without 'undue risk to the public

would also be defined within the DBE.

The postulated radioactive releases from credible 'accidents associated

with the operation of research reactors will not result in offsite radio-'

logical doses ,to'the general public eixceeding 'the Protective Action Guides

(PAGs)'of 1 rem whole body or 5 rem thyroid: Therefore,; these facilities .. would not' include the General'Emergency class of accidents requiring Federal

assistance as part of-their emergency plan.''

* Pursuant to 10 CFR 50.54(q), each licensee who is authorized-to possess

and/or 'operate a research or test reactor under a license, of the type

specified'in 10 CFR 50.21(c), 'shall follow and maintain in effect emergency

plans which meet the'requirements in:Appendix E .to 10 CFR'Part- 50..- Appendix

E to 10 CFR Part 50, 'Emergency Plans for Production and.UtilizationlFacizlities," establishes minimum requirements for emergency plans to attain'an

acceptable'state of emergency preparedness and to provide reasonable assur-

'ance'that protective measures can and will be taken to protect the health

and safety of workers and the public.

'Regulatory' Guide 2.6 (Rev.' 1,-March 1983) "Emergency Planning for

Research and Test Reactors," which is specified by Appendix E as, the guidance.to be used to determine the acceptability of research and test reactor

radiological 'emergency plans, describes a method acceptable to the NRC staff

for complying with the Commission's' emergency planning regulations. Revision.1 to Regulatory Guide 2.6 (dated March 1983), endorses American National

Standard, ANSI/ANS-15.16-1982, "Emergency Planning for Research Reactors.''

This Standard identifies'the elements oftan emergency plan which.describes

the approach to coping with emergencies 'and -minimizing-the consequences 'of

accidents at research and test reactor. facilities; The emergency plan shall

be implemented by emergency procedures.;

1American National Standard for Emergency Planning for Research Reactors','

ANSI/ANS-15.16-1982, American'Nuclear Society,-La Grange Park, IL.

1
This Standard Review Plan (SRP) has been prepared for performing

reviews and evaluations for the acceptability of research and test reactor

radiological emergency plans. The purpose of the SRP is to assure that

uniform evaluations and complete reviews are made of each research or test

reactor radiological emergency plan.

The report is organized under ten planning standards which correspond

to the guidance criteria in ANSI/ANS-15.16-1982.2

Within the research and test reactor community, the licensed thermal

power levels range from 0.1 W to 50 MW. 3 The inventory of radionuclides.

generated in reactor operations and the potential for accidents that result

in a degraded core are largely dependent upon power level and operating

history. Hence, the applicability of the planning standards to research and

test reactors is also based upon power levels. Four ranges'of power levels

(equal to or less than 100 W, greater than 100 W to less than 100 KW,

equal to or greater than 100 kW to equal to or less than 2 MW, and greater

than 2 MW) are used in the text. The applicability of the items under each

planning standard to reactors in each range is identified by an "X" in the

appropriate column of the review sections.

It should be noted that the radiation dose levels of the emergency

action levels established for the various emergency classes in Appendix I

are slightly different from-those specified for power reactors. However, in

the judgment of the NRC staff, the radiation dose'levels specified are

adequate for the credible accidents associated with the operation of research

and test reactors, and the specified action levels provide reasonable

assurance that appropriate measures associated with the action levels

specified can and-will be taken, provided appropriate emphasis is also given

to developing emergency action levels that relate directly to facility

parameters, e.g., pool water levels and area radiation monitors.

Four standard emergency classes are defined in 10 CFR 50 Appendix E.

The classes are Notification of Unusual Events, Alert, Site-Area Emergency, and General Emergency.

The General Emergency class of accidents is not credible for most

research or test reactors as this class is reserved for accidents'which

could have a significant radiological impact at substantial distances from

the reactor. Therefore, most research or test reactors would not include

this class as part of their emergency plan.

Acceptable sizes for Emergency Planning Zones-(EPZs) are given in

Appendix II as-a function of authorized steady-state thermal power level.

These are consistent with those given in ANSI/ANS-15.16-1982. The EPZ size

will be determined on a case-by-case basis for any research or test reactors

with power levels greater than 50 MW.

'The planning standards are extracted from American National Standard ANSI/

ANS-15.16-1982, with permission of the publisher, the American Nuclear

Society.

3Power level in this document means authorized steady-state thermal power

level of the reactor.

2

K-,
CONTENT OF EMERGENCY PLAN

An emergency plan shall be prepared that addresses the necessary provi- sions for coping with radiological emergencies. Activation of the emergency plan or portions thereof shall be in response to the emergency action

levels. In addition to addressing those severe emergencies that will fall within one of the standard emergency classes, the plan also shall discuss

the necessary provisions to deal with radiological emergencies of lesser

severity that can occur within the operations boundary. The emergency plan should allow for emergency personnel to deviate from actions described in

the plan for unusual or unanticipated conditions.

The plan shall consist of the following elements and address, as applicable, the provisions identified for each element.

3
AREAS OF REVIEW, PLANNING STANDARDS, AND EVALUATION ITEMS

1.0 INTRODUCTION

PLANNING STANDARD

The plan should briefly introduce the type of reactor, the reactor's

purpose, where it is'located, ahd the purposes of the emergency plan. The

:-purpose of the introduction is to provide a general orientation -and -comon

understanding about the reactor and the objective of the plan for those

members of the reactor organization, the public, and local. and federalagencies that will read and study the plan.

Applicability by Reactor

Operatinc Power Levels

Evaluation Items

>100 W to

'100 W <100 kW

k100 kW

to '2 MW >2 MW

1. The emergency plan should

include the following:

a. A description of the

reactor including authorized

power level. x x . X x

b. A description of the

location of the reactor facility

including access routes.

c. Identification of the

owner/operator.

x. x x x

x x _ _ _ x _ _ _ x

d. A definition of the

objective of the emergency plan. x x x x

5
2.0 DEFINITIONS

PLANNING STANDARD

Terms unique to the reactor facility or that have a special meaning when used in the plan should be defined in the plan.

Applicability by Reactor

Operating Power Levels

Evaluation Items

1. The emergency plan should

include definitions of words or

phrases with meanings specific or

unique to the plan or reactor.

>100 W to

M100 W <100 kW

x x_

21O0 kW

to 52 MW > MW

x x

6.
3.0 ORGANIZATION AND RESPONSIBILITIES

PLANNING STANDARD

,The'plan should describe the emergency organizationthat would be-

---activated to cope with radiological :emergencies. This includes the onsite

emergency organization and any augmentation from offsite groups.,, Persons or

groups that will fill positions in the emergency organization should be

identified by their normal everyday title. This organizational description

should include as appropriate the following evaluation items.-.

Applicability-by Reactor

Operating Power Levels

>100 W to

Evaluation Items 100 W <100 kW

210O kW

,to 2 -MW >2 MW

1. The emergency plan should

describe the following organizational considerations: 4

a. The functions as applicable to emergency planning of

Federal, State, and local

government agencies and the

assistance that they would provide in the event of an emergency.

b. The reactor's emergency

organization, including

augmentation of the reactor

staff to provide assistance for

coping with the emergency

situation, recovery from the

'emergency, and maintaining'

emergency preparedness.

c. The arrangements and

agreements, confirmed in writing

with local support organizations

that would augment and extend

the capability of the facility's

emergency organization.

'. I . . X .

x

x___x. _x x

K___ X ''' K K

fine or more of these positions may be assigned to the same incumbent.

7
Applicability by Reactor

Operating Power Levels

Evaluation Items

d. A block diagram that

illustrates the interrelationship of the facility emergency

organization to the total

emergency response effort.

Interfaces between reactor and

other onsite emergency organization groups and offsite local

support organizations and

agencies should be specified.

>100 W to

5100 W <100 kW

'10O kW

to 52 MW >2 MW

x x x x

e. The capability of the

emergency organization to

function around-the-clock for

a protracted period of time

following the initiation of

emergencies that have or could

have radiological consequences

requiring around the clock

emergency response.

f. The identification by

title of the individual in charge

of directing emergency operations,

including a line of succession,

and responsibilities and authorities and those responsibilities which may not be

delegated (such as notification

and protective action decisions). X

x

x __ x X_

g. The identification by

title of the individual, including

a line of succession, and authority and responsibilities for coordinating emergency preparedness planning, updating emergency

plans and procedures, and coordinating plans with other

applicable organizations. x x x x

h. The identification by

title of the individual, with

a line of succession, responsible

for relating information about

the emergency situation to the

news media and the public. X_ X

8
Applicability by Reactor

I. I '' Operating Power Levels

Evaluation Items , .<00 W.

i.. The identification by.

,title''of the individual, with a

.line of succession,.in-charge ..

of. radiological assessments . . including his/herresponsibilities and authority for'onsite

and offsite dose assessments and',.'

recommended protective'actions.

j. The identification'by- title of-the individual, who may authorize reentry into the

reactorbuilding orportions

of the facility that may have

been evacuated during the,

emergency. ' X

k. The identification by

title of the individual ' ..

'authorized to'terminate an.

emergency and initiate recovery actions'and be responsible for

informing the emergency organi- zation of planned organizational X.

actions or changes. . X

1. The identification by

title of the individual, who

may authorize volunteer emergency.

workers to incur radiation

exposures in excess of normal

occupational limits..

>100 W to

.clOO kW

W1OO kW

to 62 MW >2 MW

.x -

X . X

X.

__ . _

X . '',_ - . X

I . I

X - X

, . .

.. . . .

I "t .. .. ,,I -

- , I I - . - - - -

-* .

9
4.0 EMERGENCY CLASSIFICATION SYSTEM

PLANNING STANDARD

The emergency plan should describe several classes of emergency situations covering the spectrum of emergency conditions that involve the alerting

or activating of progressively larger segments of the emergency organization.

To provide for improved communications between the licensee, Federal, State

and local agencies and organizations, the most severe accidents are standardized in four classes of emergency conditions which group the accidents

according to severity of offsite radiological consequences. Each emergency

plan should include only those standard classes appropriate for dealing with

accident consequences determined to be credible for the specific facility.

Most research reactors have potential emergency situations which may occur

(e.g., personnel injury with contamination, fire, etc.) that have less

severe offsite consequences than the least severe standard class, "Notification of Unusual Events." For some research reactors, no credible accidents

are postulated which result in consequence matching the least severe class.

However, planning for onsite emergencies is important. Preparedness for the

onsite emergencies should be accomplished by identifying them and including

in the plan those elements of this standard commensurate with the postulated

emergency situations.

Each class of emergency should be associated with particular emergency

action levels and with particular immediate actions to provide appropriate

graded response. In order of increasing severity, the four standard emergency classes are described in qualitative terms in the following subsections:

4.1 NOTIFICATION OF UNUSUAL EVENTS. Notification of unusal events may

be initiated by either man-made events or natural phenomena that can be

recognized as creating a significant hazard potential that was previously

nonexistent. There is usually-time available to take precautionary and

corrective steps to prevent the escalation of the accident or to mitigate

the consequences should it occur. No releases of radioactive material

requiring offsite responses are expected.

One or more elements of the emergency organization are likely to be

activated or notified to increase the state of readiness as warranted by

the circumstances.

Although the situation may not have caused damage to the reactor, it

may warrant an immediate shutdown of the reactor or interruption of nonessential routine functions.

Situations that may lead to this class include: (1) threats to or

breaches of security, such as bomb threats or civil disturbances directed

toward the reactor; (2) natural phenomena, such as tornados in the immediate

vicinity of the reactor, hurricanes, or earthquakes felt in the facility;

(3) facility emergencies, such as prolonged fires, fuel damage indicated by

high coolant fission product activity, or high offgas activity.

10

K-,
4.2 ALERT. Events leading to an alert would be of such radiological

significance as to require notificationET-the emergency organization and

its response as appropriate for the specific emergency'situation. Under

this class it is unlikely that offsite response or monitoring would be

necessary. Substantial modification of reactor operating status is a highly

probable corrective action. Protective evacuations or isolation of certain

areas within the operations boundary or within the site boundary may be

'necessary. Situations that may lead to-this'class include: (1) severe

failure of fuel cladding-or-of.fueled-experiments where containment boun-.'

daries'exist to reduce releases or less severe'cladding failures in situations where fission products are not well contained,'or (2)'significant,

releases of radioactive materials as a result of experiment failures. -

-4.3 !SITE AREAEMERGENCY.- A site area emer ency may'be initiatedwhen

events such asmajor damage of fuel7o-Fcladding and actual or imminent . "failure of other physical barriers-containing fission-products in reactor

fuel or fueled experiments have'occurred..and projected offsite radiological

consequenc'es exceed Appendix I action levels. Monitoring at-the'site..

"'boundary should be conducted to assess the'need for offsite protective

actions. Protective measures on site may. be necessary. -

4.4 GENERAL EMERGENCY. A general emergency may be initiated by

accidents which result -in an uncontrolled release of radioactive material

into the air, water, or ground to the extent that protective actions offsite

may be necessary. This class of accident is not credible for most research

reactors. -Therefore, most'research reactors would not include this class as

part of their emergency plans.

A protective action that may be recommended to offsite authorities may

be to shelter the general public within the EPZ. State and local government

response organizations have the ultimate responsibility for initiating and

implementing any recommended offsite protective actions.

Applicability by Reactor."

Operating Power Levels'

>100 W to .. 100 kW.

Evaluation Items i100 W <100 kW to 2 MW .>2 HW

1. The emergency plan should

contain:

a. An emergency classification

system consistent with the

planning standard.

b. In an Appendix to the plan,

a listing by title of implementing procedures for each

class of emergency.

K X X X

X X X X

11
I

5.0 EMERGENCY ACTION LEVELS

PLANNING STANDARD

Because of the wide-diversity in research reactors (power level,

engineered safety features, site environment, etc.), those conditions which

might initiate or signal a radiological incident having particular offsite

consequences will vary widely among facilities. Action levels may be

specified for effluent monitors or other plant parameters for which the dose

rates and radiological effluent releases at the site boundary can be projected. To establish effluent action levels, facilities that have meteorological information available may base the action levels on actual meteoirological conditions; otherwise, the criteria for downwind concentration,

Section 4 of ANSI/ANS 15.7-1977, "Research Reactor Site Evaluation," should

be used. Each-emergency plan should establish emergency action levels

appropriate .for the specific facility and consistent with Appendix I. The

emergency plan should include emergency action levels to initiate protective

actions for members of the general public onsite. The protective action

guide (PAG) shall be 1 rem whole body or 5 rem thyroid.

Applicability by Reactor

ODeratina Power Levels

'-

, ._ ........ , ..................................

.._ .

Evaluation Items

1. Each licensee's emergency

plan should contain:

>100 W to

<100 W <100 kW

'100 kW

to '2 MW >2 MW

a. Emergency action levels

which are appropriate to the

specific facility and consistent

with Appendix I. To the extent

possible specify effluent

monitors used to project dose

rates and radiological effluent

releases at the site boundary. X X X X

______________

5Manual of protective Action Guides and Protective Actions for Nuclear Incidents,

EPA-520/1-75001, Sept. 1975, U.S. Environmental Protection Agency.

12
6.0 EMERGENCY PLANNING ZONES

PLANNING STANDARD

-. As part'of emergency planning, the, reactor.owner/operator of a facility

that identifies radiological emergencies which result in offsite plume exposures exceeding 1 rem whole body or 5 rem thyroid should identify an'

emergency planning zone (EPZ). The postulated radioactive release from -

credible accidents provides the basis for determining the need for an EPZ.

The size -of the'EPZ should be established so that the dose to individuals

beyond the EPZ is not projected to exceed the PAG. As an alternative to

performing such calculations, the EPZ sizes in Appendix II may be adopted

according to the power level. . .

Applicability by Reactor

I ' Operating Power'.Levels

Evaluation Items

1. Ensure that the emergency

plan identifies the EPZ.

2. If the EPZ is not consistent

with Appendix II, the plan shall

include an acceptable 'basis for'--

the EPZ.

>100 W -to'

<100 W <100 kW

x . _. _ , .

*100 kW

to 52 MW' >2 MW

' X ' i X ,x

.- - _ -

.X .X X X

, fI

... , . t . . . .I . . . ... .1 ..

I ..... I

0, . - ;-- I ;' I

I . :1

13
a7.0 EMERGENCY RESPONSE

PLANNING STANDARD

Emergency response measures should be identified for each emergency.

These response measures should be related to the emergency class and action

levels that specify what measures are to be implemented.

Applicability by Reactor

Operating Power Levels

Evaluation Items

1. The emergency plan should

cover the following notification

information for emergency

response:

>100 W to

'100 W <100 kW

'100 kW

to <2 MW : >2 MW

a. The actions to notify and

mobilize the emergency organization and the applicable oftsite

support organizations for each

emergency class.

b. The location(s) of current

notification lists.

X X X_. X

X X X X

c. Describe the contents of

initial and followup emergency

messages to the NRC and, when

applicable, to offsite authorities. To the extent known,

these messages should include

the following:

(1) Name, title and telephone number of caller, and the

location of the incident and the

emergency class. X

(2) Description of emergency

event. X

X

x

x

(3) Date and time of incident

initiation. X

(4) The type and quantity of

radionuclides released or expected

to be released. X X

14

K>
Applicability by Reactor

Operating Power Levels

--Evaluation- Items

(5) Impact of releases

and recommended offsite emergency

actions.

>100 W to

i10O W <100 kW

>100 kW

t o S !2 MW>2 MW

. .. I , .r.- x

d. A method is established

to insure that offsite authorities

have received the initial message

and that it is authentic. -- _ _X X ,I _ :. - -- - -- "

2. The emergency plan should

cover the following assessment

considerations:

a. A description of methods

for gathering and processing

information for assessment actions. x X

3. The emergency plan should

provide a summary description

of those actions that could be

taken to mitigate or correct

the problem for each emergency

class. x x _ _ _ x _ _ _ x _ _

4. The emergency plan should

describe protective actions

appropriate for the emergency

class. The emergency plan

should include the following:

a. Conditions for either

partial or complete onsite

evacuation, evacuation routes,

and primary and alternate

assembly areas. x _ _ _ x _ _ x _ _ _ x _ _

b. Methods to ensure personnel accountability and the

segregation of potentially contaminated personnel.

c. Protective measures and

exposure guidelines for emergency personnel.

x x _ _ _ x _ _ _ x _ _

x _ _ _ x _ _ _ x _ _ _ x _ _

15
__

Applicability by Reactor

Operating Power Levels

Evaluation Items

d. Provisions for isolation

and access control of facility

areas to minimize exposures to

radiation and the spread of

radioactive contamination.

>100 W to

0100 W <100 kW

'10O kW

to 62 MW i >2 MW

x _ _ _ x _ _ _ x _ _ _ x _ _

e. The methods for monitoring radiation dose rates and contamination levels, both onsite and

offsite, including provisions

for transmitting collected information and data to the element of the emergency organization responsible for accident

assessment. x _ _ _ _ _ _ x _ _ _ x _ _

16
8.0 EMERGENCY FACILITIES AND EQUIPMENT

- PLANNING STANDARD

The emergency plan should briefly describe the emergency facilities, types

of equipment'and their location.

Applicability by Reactor

Applicabito b Ro eactor

Operating Power Levels -

Evaluation Items

1. The emergency plan should

describe an emergency support

center (ESC).

-0 >100WW to

_100 W ' <100 kit

X, .

to <2 MW >2 MW

x .___

2. Representative types of

monitoring and sampling equipment to be used for accident - * assessment and their-location.

These should include:

e a. Portable and fixed radiological-monitors. -

b. Sampling equipment.

c. Instrumentation for spec- ific radionuclide identification

and analysis.'

d. Personnel monitoring equipment.

e. The plan should also describe nonradiological monitors

or indicators that may provide

pertinent'information; for

example:

(1) Reactor instrumentation.

X___ X_

X I.'.' X '-

X

I'X -

-

X

X_

., I

X

X - . . .. , ......

: - - - -- -,: , x___

, . . * . .

v , I. .

I : . . . (2) V Y Fire detectors,

earthquake sensors, etc. X__ X

(3) Source of meteorological data representative

of facility location. X

17
Applicability by Reactor

Operating Power Levels

Evaluation Items

3. The emergency plan should

identify those measures that

will be used to provide necessary

assistance to persons injured or

exposed to radiation. The capabilities for decontamination, administering first aid, transporting

injured personnel, and arrangements for medical treatment should

be described. The following items

should be included:

>100 W to

S10O W <100 kW

210O kW

to 52 MW >2 MW

a. Facilities for personnel

decontamination.

b. Methods for handling and

transporting contaminated injured

personnel.

c. Written agreements with

hospitals to ensure that medical

services are available and the

staff is prepared to handle

radiological emergencies.

4. The emergency plan should

adequately identify the emergency

communications systems that

will be available to communicate

instructions and information

both onsite and offsite throughout the course of an emergency.

X X X X

XI X _ X X

X X X X

X_ X X X

5. Facilities planning for a site

area emergency should establish

reliable means of cotmunication,

e.g., public telephone and radio,

that is compatible with local offsite support groups. X

18
9.0 RECOVERY

PLANNING STANDARD

-This element of'the emergency planshould describe the.criteria for

restoring the reactor facility to a safestatus-including reentry into the

reactor building or portions of the facility-that-may have been evacuated

because of the accident. The operations to recover from most severe accidents

will be complex and depend on the actualvconditions at the-facility. It is not

practicable to plan detailed recovery actions'for all conceivable situations.

Applicability by Reactor

Operating Power Levels

Evaluation Items

1. The emergency plan should

specify:

>100 W to 210O kW

'100 W clO kW to 2V MW- ->2 MW

a.

will be

needed.

That recovery procedure(s)

written and approved as . I

.. .. . . ~ X 4~.1X .:<. X

.;19 .
10.0 MAINTAINING EMERGENCY PREPAREDNESS

PLANNING STANDARD

The emergency plan should describe the elements necessary for maintaining

an acceptable state of emergency preparedness. A description should be provided of how the effectiveness of the emergency plan will be maintained,

including training, review and update of the emergency plan and associated

implementing procedures, and maintenance and inventory of equipment and

supplies that would be used in emergencies.

Applicability by Reactor

Operatina Power Levels

' i

Evaluation Items

>100 W to

0100 W <100 kW

'-100 kW

to 52 MW >2 MW

1. The emergency plan should

describe an initial training

and periodic retraining program

designed to maintain the ability

of emergency response personnel

to perform assigned functions for

the following:

a. Personnel responsible for

decisionmaking and transmitting

emergency information and instructions. x x x x

b. Personnel responsible for

accident assessment.

c. Radiological monitoring

and analysis teams.

d. First aid and rescue

personnel.

e. Medical support personnel

f. Police, security, ambulance and fire fighting personnel.

x x x x

x x x x

x x_ x x

x

x_

x x X_

x x x

2. The emergency plan should

provide for:

20
Applicability by Reactor

Operating Power Levels

Evaluation Items

>100 W to

$100 W <100 kW

k100 kW

to ;52 MW I>2 MW

a. Annual onsite emergency

drills, to be conducted as

action drills.6

b. Provision for critiques of

all drills, including timely eValuation of observer comments and

correction of identified deficiencies.

c. -Development of-written

scenarios for conducting annual

action drills.

3. The emergency plan should

provide for a biennial review

and update of the emergency -

plan and-implementing procedures ---

and agreements with offsite

support organizations and agencies

including:

X X - - -X -. X . . ..

1 . .

X' X I - ; - XX X - , X.

X

X.

: '. I . '..' -

' I . I I .

a. Reviews and approvals by

those responsible for emergency

planning.

b. Incorporation of modifications resulting from action

drills or changes in the .

facility or environs.

X XX X_X X

X Xc. Timely forwarding of

approved amendments to the plane

agreements, and implementing

procedures to authorized

individuals, agencies and

support organizations. X X X_ X

bAn action drill tests the integrated capability of the emergency plan, or a

component thereof, and may include instruction periods to develop and maintain

skills in a particular operation.

21
Applicability by Reactor

Operating Power Levels

Evaluation Items

4. The emergency plan should

describe the provisions to

ensure operational readiness

of emergency communications

and emergency health physics

equipment by including:

>100 W to

S100 W <100 kW

>100 kW

to 62 MW >2 MW

a. Required maintenance

and minimum calibration

frequency.

b. Functional testing including minimum frequency.

x x___ x_ _

x___ x__ x__

x

x

c. Minimum frequency of.

inventory for equipment and

supplies. x _ _ _ x _ _ x _ _ _ x _ _

22

K-I
APPENDIX I

EMERGENCY CLASSES

Emergency Class Action Level' Purpose

Notification of

Unusual Events

Alert

Actual or projected radiological

effluents at the site boundary

exceeding 10 MPC2 when averaged

over 24 hours, or 15 mrem whole

body accumulated in 24 hours

Report or observation of severe

natural phenomenon

Receipt of bomb threat

Actual or projected radiological

effluents at the site boundary

exceeding 50 MPC 2 when averaged

over 24 hours, or 75 mrem whole

body accumulated in 24 hours

Actual or projected radiation

levels at the site boundary of

20 mrem/hr for 1 hour whole body

or 100 orem thyroid dose

Actual or projected radiological

effluents at site boundary exceed

ing 250 MPC2 when averaged over

24 hours, or 375 frem whole

body accumulated in 24 hours

Actual or projected radiation

levels at the site boundary of

100 mrea/hr for 1 hour whole

body or 500 mrem thyroid dose.

Site Area

Emergency

(1) Ensure that the first

step in any response later

found to be necessary has

been carried out. (2) bring

the operating staff tp a state of readiness, and

(3) provide systematic

handling of unusual events

information and decisionmaking

(1) Ensure that emergency

personnel are readily

available to respond if

the situation becomes more serious or to perform

confirmatory radiation

monitoring if required,

and (2) provide current

offsite authorities

status information

(1) Ensure that response - centers are manned, (2) ensure that monitoring

teams are dispatched,

(3) ensure that personnel

required for evacuation of

onsite areas are at duty

stations, (4) provide con- sultation with offsite

authorities and (5) provide

information for the public

through offsite authorities

(1) Initiate predetermined

protective actions for

the public, (2) provide

continuous assessment of

information from licensee

and offsite organization

measurements, (3) initiate

additional measures as indicated by actual or potential releases, (4)

provide consultation with

offsite authorities, and

(5) provide updates for the

public through offsite

authorities.

General Emergency Sustained actual or projected

radiation levels at the site

boundary or 500 mrem/hr whole

body.

Actual or projected dose at the

site boundary in the plume

exposure pathway of 1 rem whole

body or 5 rem thyroid

'The situations that may lead to an emergency class described in the subsections

of Section 4.0 may be referenced as emergency actions levels appropriate to the

emergency class

2Maximum Permissible Concentration (MPC) as listed in Title 10, of the Code of

Federal Regulations, Part 20, 'Standards for Protection Against Radiation,'

Appendix B. Table II Column I

23
APPENDIX II

Alternate Method For Determining

The Size of an EMERGENCY PLANNING ZONE'

Authorized Power Level Acceptable EPZ Size

'2 MW Operations boundary

>2 MW and 510 MW 100 meters

>10 MW and 520 MW 400 meters

>20 MW and <50 MW 800 Meters

>50 MW Will be determined on a case-by-case

basis

'Calculations are based on:

D. Bruce Turner, Work Book of Atmospheric Dispension

Air Programs. U.S. Environmental Protection Agency,

(1970)

Estimates, Office of

Washington, D.C.

D. H. Slade, Ed., "Meteorology and Atomic Energy." U.S. Atomic Energy

Commission, Washington, D.C. (1968); and

WASH 1400 (NUREG 75/014), "Reactor Safety Study," Appendix VI. U.S. Nuclear

Regulatory Commission, Washington, D.C. (1975).

25
I~ NRC FORtM33S go Ill

US NUCLEAR REGULATORY COMMISSION

BIBLIOGRAPHIC DATA SHEET

I REPORTNUM8ER AU~ssrior DOCJ

NUREG-0849 It--l-1011 1. - -i TITLE AND SUBTITLE SAd VMumes No. aoparensjw 2 kLeae bltavJ

Standard Review Plan for the Review and Evaluation of

Emergency Plans for Research and Test Reactors 3 RECIPIENT S ACCESSION NO

7 AUTHORISI S DATE REPORT COMPLETED

MONTH YEAR

September 1983

9 PERFORMING ORGANIZATION NAME AND MAILING ADDRESS fincha- Zip COel DATE REPORT ISSUED

Division of Emergency Preparedness -and Engineering Respons 0 To er IYE"583

Office of Inspection and Enforcement

U.S. Nuclear Regulatory Commission 6 ALE*~tar k

Washington, DC 20555 a e blki

12 SPONSORING ORGANIZATION NAME AND MAILING ADDRESS fliclume Zp Codel . t10 PROJECT TASK. WORK UNIT NO

Same as 9,'above. II FINNO

13 TYPE OF REPORT PERIOO COVEPED IinC"IeuI. od"SA

Technical Report

15 SUPPLEMENTARY NOTES 14 teLo-e Oriwel

E

16 ABSTRACT t7000 worr 0s'le

This document provides a Standard Review Plan to assure that complete and uniform

reviews are made of research and test reactor radiological emergency plans. The report

is organized under ten planning standards which correspond to the guidance criteria in

American National Standard ANSI/ANS 15.16 - 1982 as endorsed by Revision 1 to Regulatory

Guide.2.6. The applicability of the items under each planning standard is indicated by

subdivisions of the steady-state thermal power levels at whicb the reactors are licensed

to operate. Standard emergency classes and example.action levels for research and test

reactors which should initiate these classes are given in an Appendix.

The involvement of the Federal Emergency Management Agency IFEMA) concerning

emergency response of research and test reactors is not required. However, the NRC will

request FEMA assistance in the review process if the NRC staff identifies a facility

which postulates a credible accident that.could result in offsite radiological doses to

the general public exceeding the Protective Action Guides of 1 rem whole body or 5 rem

thyroid.

17 KEY WORDS AND DOCUMENT ANALYSIS 17j DESCRIPTORS

17b IDENTIFIERS OPEN ENDED TERMS

I - __ __ __ - . ___- . ill - _e - -CC AVAILABILITY STATEMENT I", -1dh0V9tff ftu -I -,"I" - - -- .

I lInlimitomd

* .

2C SEiCUJITY CL4.SS ITA-1 P&9rf 122 PRICE

I I - - ..- I

NRC FORM 33S '1Is",
Appendix 12.3

Regulatory Guide 5.59, Revision 1

Standard Format and Content

for a Licensee Physical Security Plan

for the Protection of Special Nuclear Material

of Moderate or Low Strategic Significance

ST~,ommm FORMAT Am ComnTm
0 R03t REG Revision I

U.S. NUCLEAR REGULATORY COMMISSION February 198

:" REGULATORY GUIDE , . OFFICE OF NUCLEAR REGULATORY RESEARCH

REGULATORY GUIDE 5.59

(Task SG 229.4)

STANDARD FORMAT AND CONTENT

FOR A LICENSEE PHYSICAL SECURITY PLAN

FOR THE PROTECTION OF SPECIAL NUCLEAR MATERIAL

OF MODERATE OR LOW STRATEGIC SIGNIFICANCE

UtSNRC REGULATORY GUIDES

R gutatOey Ou1dos are Iased to 0escrlos and Make available to the public mothos acceptable to the NRC Staff of Implementing soecific Parts of the Commlslon s regulations, to delineatS tech- nowis used by the etaff In evaluating specIfic problems or postu- atted accIdent Or to provide guIdance to appllents Regulatory Guides are no( *ubstItute for regulatIons, and ComplIanc Veth them Is not required. Methods and Solutions different tfrom those set out In the guIdes wIll be acceptable It they provIde a basis for the findings requlsite to the Isuance or continuance of a permit or license by the Commirslon.

This gude- was Issued after consideratlon of comments received from the public. Comments and sugestions for Improvements In theo guides are encouraged at all times, end guides will be rievised as acoroorlete. to accommodate Comments sndl to reflect new Informa- tion or experience.

Comments Should be sent to the Secretary of the CommIsslon. U S. Nuclear Regulatory Commisso. Washington. D.C. 20555. Attention: DockSting end 5evIce Branch.

The guides are Issued In the following tsn broad dirislons.

I Power Reactors 6. Products 2. Research and Tert Reactors 7. Transportation 3. Fuels and Materials Facilities B. Occupational Health 4 Environmental end SIting 9. Antitrust and Financial Review 5. Materials and PtantoPotectIon 10. General

CooiS of Issued guides may be purchased at the current Government

Printing Office Price. A subscription service for future guidls in so ciiC divisiSons Is available throu91 the Government Printing Office. Information on tne subscriotion service and current GPO prices may be Obtained bY writing the U.S. Nuclear Regulatory CommIssion.

washington. D C 20555. Attention Publications Sales Manager
TABLE OF CONTENTS

Page

INTRODUCTION.........................................................

PART I SPECIAL NUCLEAR MATERIAL'OF MODERATE STRATEGIC -

- SIGNIFICANCE ...................................................

, ' Chapter 1' USE'AND STORAGE AREA AT A FIXED SITE.................

1:1'Area'Where Material 'Is Used.'

1.2 Area Where Material Is Stored.

Chapter 2 DETECTION DEVICES AND PROCEDURES AT A FIXED SITE..

vii

5.59-1

5.59-3

5.59-3

5.59-6

5.59-9

-*2.1-Earliness of Detection ................................ 5.59-9

2.2 Detection Through Monitoring Controlled Access Areas 5.59-10

Chapter 3 ACCESS CONTROL AT A FIXED SITE ..............................59-

3.1 Preauthorization Screening .5.59-12

3.2 Badging System.:'................... 5.59-13

3 3 -Lock System...................................5......... 5.59-13

3.4 Personnel Entry Control System .......................... 5.59-14

3;5 Escort System .........I.............................. 5.59-15

3.6 3.6 Search..5.59-15 Serc . ....................................................... .... 5.591

Chapter 4 SECURITY ORGANIZATION AT A FIXED SITE . . ..................... 5.59-17

Chapter 5 COMMUNICATIONS AT A FIXED SITE .............................. >-5.59-18

Chapter 6 RESPONSE PROCEDURES AT A FIXED SITE ......................... 5.59-19

Chapter 7 MATERIAL TRANSPORTATION REQUIREMENTS ...... ................. 5.59-20

7.1 Advance Notification to Receiver ......................... 5.59-20

7.2 Receiver Confirmation .................................. 5.59-21

7.3 Inspection....-.,_-., 5.59-21

7.4 Responsibility for In-Transit Physical Protection n 5.59-21

:-'Chapter 8 RECEIVER REQUIREMENTS--TRANSPORTATION. ....................... 59-22

8.1 Inspection ............-.... ........ ' .. 5.59-22

8.2 Notification to Shipper ................................ ..5.59-22

8.3 Responsibility for In-Transit Physical Protection..... 5.59-22

,. . .

iii
TABLE OF CONTENTS (Continued)

Paae

Chapter 9 IN-TRANSIT PHYSICAL PROTECTION REQUIREMENTS ................ 5.59-24

9.1 Communications... ....................... 5.59-24

9.2 Minimum Transit Times .................................. 5.59-26

9.3 Preauthorization Screening ............................. 5.59-26

9.4 Response Procedures .................................... 5.59-27

9.5 Notification ................ 5.59-27

9.6 Lost Material Notification ............................. 5.59-27

9.7 NRC Notification ............... 5.59-28

Chapter 10 TRANSFER AND CONTROL REQUIREMENTS ......................... 5.59-29

10.1 Carrier Transfers ................................... 5.59-29

10.2 Control of Shipments ................................ 5.59-30

Chapter 11 EXPORT REQUIREMENTS ........................................ 5. 59-32

Chapter 12 IMPORT REQUIREMENTS .5.59-33

12.1 Security Requirements............................... 5.59-33

12.2 Notification .5.59-33

Chapter 13 ORDERS TO DELAY SSNM SHIPMENTS .5.59-34

.13.1 Notification of Receipt of SSNM .5.59-36

PART II SPECIAL NUCLEAR MATERIAL OF LOW STRATEGIC SIGNIFICANCE . . 5.59-37

Chapter 1 USE AND STORAGE AREA AT A FIXED SITE ....................... 5.59-38

1.1 Areas for Use and Temporary Storage .................... 5.59-38

1.2 Areas for Permanent Storage ............................ 5.59-40

Chapter 2 DETECTION DEVICES AND PROCEDURES AT A FIXED SITE ........... 5.59^43

2.1 Earliness of Detection ................................. 5.59-43

2.2 Monitoring Controlled Access Areas ..................... 5.59-44

Chapter 3 SECURITY RESPONSE AT A FIXED SITE .......................... 5.59-49

Chapter 4 RESPONSE PROCEDURES AT A FIXED SITE ........................ 5.59-50

Chapter 5 MATERIAL TRANSPORTATION REQUIREMENTS ....................... 5.59-51

5.1 Advance Notification ................................... 5.59-51

5.2 Receiver Confirmation .................................. 5.59-52

5.3 Container .............................................. 5.59-52

5.4 Inspection ............................................. 5.59-52

5.5 Responsibility for In-Transit Physical Protection ...... 5.59-53

iv
TABLE OF CONTENTS (Continued)

Page

Chapter 6 RECEIVER REQUIREMENTS--TRANSPORTATION ...................... 5.59-54

6.1 Inspection ............................................. 5.59-54

6.2 Notification ........................................... 5.59-54

6.3 Responsibility for In-Transit Physical Protection ...... 5.59-54

Chapter 7 IN-TRANSIT PHYSICAL PROTECTION REQUIREMENTS ................ 5.59-56

7.1 Response Procedures .................................... 5.59-56

7.2 Notification ........................................... 5.59-56

7.3 Lost Material Notification ............................. 5.59-57

Chapter 8 EXPORT REQUIREMENTS ...................... 5.59-58

Chapter 9 IMPORT REQUIREMENTS ...................... 5.59-59

9.1 Security Requirements .................................. 5.59-59

9.2 Notification ............................................ 5.59-59

VALUE/IMPACT STATEMENT .................. 5.59-60

v
INTRODUCTION

The Atomic Energy Act of 1954, as amended, directed the U.S. Atomic Energy

Commission (AEC) to regulate the receipt, manufacture, production, transfer,

possession, use, import, and export of special nuclear material (SHM) in order

to protect the public health and safety and to provide for the common defense

and security. *The Energy Reorganization'Act of 1974 transferred all the

licensing and related regulatory functions of the AEC to the Nuclear Regulatory

Commission (NRC).

The principal requirements with' respect to the physical protection of

licensed activities against industrial sabotage and with respect to the physical.protection of special nuclear material in transit are found in 10 CFR

Part 50, "Domestic Licensing of Production and Utilization Facilities"; Part

70, 'Domestic Licensing of Special Nuclear Material"; Part 73,, "Physical

Protection-of Plants and Materials"; and Part 1o1, "Export and Import of

Nuclear Equipment and Materials." - -. -

Paragraph 50.34(c) of 10 CFR Part 50 and paragraphs 70.22(g), 70.22(h),

and 70.22(k) of 10 CFR Part 70 identify the physical protection information

that must be provided in a physical security planas part of 'a license application. This planis required in order for the applicant'to demonstrate -

compliance with the specific physical protection requirements of 10 CFR Part 73

and must be submitted with each application for -a license to possess or use SNM

(or for a license authorizing transport oridelivery of SNM), :except.fora

license to possess, use, or transport less than 10 kg of SNM of low-strategic

significance, in which case a physical security plan is not required. However,

for the protection of SNM of low-strategic-significance, the licensee is

required to meet the'requirements of § 73.67, 'Licensee Fixed Site and

In-Transit Requirements for the Physical Protection of -Special Nuclear Material

of Moderate and Low Strategic Significance," of 10 CFR Part.73.

Any guidance in this document related to information collection activities

has been cleared under OMB Clearance Nos.' 3150-0002, 3150-0009, and 3150-0011.

This regulatory guide describes the information required in the physical

security plan submitted as part of an application for a license to possess,

use, or transport SNM of moderate strategic significance or 10 kg or more of

SNl of low strategic significance and recommends a standard format for-

*presenting the information in an orderly arrangement. This standard format

will thus serve as an aid to uniformity and completeness in the preparation and

review of the physical security plan of the license application. This document

can also be used as guidance by licensees possessing or transporting less than

10 kg of SNM of low strategic significance in understanding -the intent and

implementing the requirements of paragraphs 73.67(a), 73.67(f), and 73.67(g) of

10 CFR Part 73.

Aside from providing guidance for the standard format and content ofphysical security plans, this regulatory guide explains the intent of the

various provisions of the regulation. The intent of each.requirement is found

in the discussion of each subsection and is implicitly provided by outlining

alternative systems that could be used to fulfill-the.requirements.' The

vii
discussion section and list of alternatives should provide the licensee with

the sense of the NRC regulations.

This guide is divided into two parts. Part I, "Special Nuclear Material

of Moderate Strategic Significance," provides a standard format for preparing

the licensee's security plans and provides guidance to licensees who possess,

use, or transport SNM of moderate strategic significance. Chapters 1 through 6

of Part I apply to applications for a license to possess or use at any fixed

site, or at contiguous sites subject to control by the licensee, SNM of

moderate strategic significance. Chapters 7 through 13 of Part I apply to

applications for authorization to transport or deliver to a carrier for

transport SNM of moderate strategic significance.

Part II, "Special Nuclear Material of Low Strategic Significance," provides a standard format for preparing the licensee's security plan for

licensees who possess, use, or transport more than 10 kg of-SHM of low

strategic significance. It also provides guidance to all licensees who

possess, use, or transport SNM of low strategic significance. Chapters 1

through 4 of Part II apply to applications for a license to possess or use at

any fixed site, or at contiguous sites subject to control by the licensee, more

than 10 kg of SNM of low strategic significance. Included in this category are

licensees who have nuclear power reactors under construction and are seeking a

license to possess nuclear fuel onsite prior to obtaining their operating

license. Chapters 5 through 9 of Part II apply to applications for

authorization to transport or deliver to a carrier for transport more than

10 kg of SNM of low strategic significance.

Table 1 shows the type and amount of SNM covered in § 73.67 of 10 CFR

Part 73. It should be-noted, as stated in the footnote to Table 1, that (1)

plutonium with an isotopic concentration exceeding 80 percent or more in

Pu-238, (2) special nuclear material that is not readily separable from other

radioactive material and that has a total external radiation dose rate in

excess of 100 rems per hour at a distance of 3 feet-from any accessible surface

without intervening shielding, and (3) sealed plutonium-beryllium neutron

sources totaling 500 grams or less of contained plutonium at any one site or

contiguous sites are exempt from the requirements of S 73.67 of 10 CFR Part 73.

This guide has been prepared to minimize time lost because of incomplete

physical security plans and to standardize the review process. Applicants are

encouraged to-prepare their physical security plans in accordance with this

guide and to provide information in each section to support the conclusion that

they will be able to operate in accordance with the pertinent regulations.

Although conformance with this guide is not required, the format and content

presented are acceptable to the NRC staff.

As developments and changes in the nuclear industry occur, the Commission's requirements for information may need modification; revisions to this

guide will be made as necessary to accommodate these changes.

Purpose and Applicability

This standard format has been prepared as an aid to uniformity and completeness in the preparation and review of the physical protection section of

license applications and to clarify the intent of the regulations. The

viii
( (

Table 1

'- .CATEGORIES OF SPECIAL NUCLEAR MATERIAL

MODERATE STRATEGIC - LOg STRATEGIC

MATERIAL*` ENRICHMENT SIGNIFICANCE SIGNIFICANCE

: 1. Plutonium Less than 2,000 g but more

than 500 g .

2. tUranium-235

-. .

20% or more in'

U-235 isotope

10% or more but

less than 20%

in U-235 isotope

Above natural - but less,-than

10%

Less than 5,000 g but more

than 1,000 g

10,000 g or more500 g or less but more

than 15'g

1,000 g or less but

more than 15'g

Less than 10,000 "gbut

more than 1,000 g

I.

x

3. - Uranium-233

' 4. Uranlum-235,

uranium-233,

* and plutonium i in com-.

.' :-' bination.:- I . I :.

. . .

U-23

enri

or ff

* ' '-: Less than 2,000 g but more . than500 g

35 portion Less than 5,000 g according

ched to 20% to the formula: grams,=

nore. . (grams contained U-235) +

2.5 (grams U-233,+ grams.-

plutonium) but more than

formula: grams = (grams

U-235) + 2.0 (grams'U-233

+.grams plutonium)

10,000 g or more.''

500 g-or less but more

than 15 g

1,000 g or less according

to the-formula: grams = (grams contained U-235) +

2.0'(grams U-2334+ grams plutonium) but more

than 15 g according tothe formula: grams =

grams contained U-235 +

grams U-233 + grams *plutonium. -:- .. - -

. -

t

.j

*The following materials are exempt:.

1. Special nuclear material that is not readily separable-from the radioactive material and

that has a total external radiation dose rate in excess of 100 reis per hour at a distance

of.3 feet from' any accessible surface without intervening shielding,

2. Plutonium 'with an isotopic concentration of 80 percent or more in Pu-238, and

3. Sealed'plutonium-beryllium neutron sources totaling 5OO grams or less of contained-plutonium

at anyone site or' contiguous sites. .
information this guide contains will help the licensee plan a physical protection system designed to detect the theft of SHN of moderate or low strategic

significance. The physical protection subsystems identified are those that

will usually be included in a protection system that would normally be capable

of meeting the performance requirements of paragraph 73.67(a) of 10 CFR

Part 73. However, it is recognized that at any particular site there may be

some subsystems and components not' needed or additional-ones needed to meet

these performance requirements. In these cases, applicants are encouraged to

address in the license application specific departures of subsystems or

components from this guide.

The information requested in this guide is the minimum needed for the

review of a physical security plan. Additional information may be required for

completing the staff review of a particular plan and should be included as

appropriate. It: is also the applicant's responsibility-to be aware of new and

revised NRC regulations. The information provided should be up to date with

respect to the state of technology for the physical protection techniques and

systems that the applicant proposes to use.

In cases where NRC-approved security plans are already in existence and

where the measures included.therein meet or exceed the requirements of § 73.67

of 10 CFR Part 73, licensees may reference in their proposed security plans

those sections of the NRC-approved security plans that are applicable.

Information and procedures delineated in the regulatory guides in Division 5, "Materials and Plant Protection," that are appropriate to certain sections of the physical security plan may be incorporated by reference.

Applicants should discuss their plans and programs with the NRC staff

before preparing the applications. These discussions should give particular

emphasis to the depth of information required for the plans.

Upon receipt of an application, the NRC staff will perform a preliminary

review to determine whether the application provides a reasonably complete

presentation of the information needed to form a basis for the findings

required before issuance of a license. The standard format will be used by the

staff as a guideline for identifying the type of information needed. If an

application does not provide a reasonably complete presentation of -the

necessary information, further review of an application will be suspended until

this needed information is provided.

Use of Standard Format

The applicant should follow the numbering system of the Standard Format

down to the level of section (e.g.., 3.4). Under some circumstances, certain

sections may not be applicable to a specific application. If so, this should

be clearly stated and sufficient information should be provided to support that

conclusion.

The applicant may wish to submit in support of the application information

that is not required by regulations and is not essential to the description of

the applicant's physical protection program. Such information could include,

for example, historical data submitted in demonstration of certain criteria,

discussion of alternatives considered by the applicant, or supplementary data

x
regarding assumed models, data, or calculations. This information should be

provided as an appendix to the application.

Upon completion of the application, the applicant should use the Table of

Contents of the Standard Format as a checklist to ensure that each subject has

been addressed.

Style and Composition

A table of contents should be included in each submittal.

The applicant should strive for clear, concise presentation of information. Confusing or ambiguous statements and general statements of intent

should be avoided. Definitions and abbreviations should be consistent throughout the submittal and consistent with generally accepted usage.

- Wherever possible, duplication of information should be avoided. Thus,

information already included in other sections-of the 'applications may be

covered by specific reference to those sections.

Where numerical values are stated, the number of significant figures

should reflect the accuracy or precision to which the number is known. The use

of relative values should be clearly indicated.

Drawings, diagrams, and tables should be used when information may be

presented more adequately or conveniently by such means. These illustrations

should be located in the section where they are first referenced. Care should

be taken to ensure that all information presented 'in drawings is legible,-that

-symbols are defined, and that drawings are not reduced to the extent that they

cannot be read by unaided normal eyes.

Physical Specifications of Submittals

All material submitted in an application should conform to the following

physical dimensions of page size, quality of paper and inks, numbering of

pages, etc.: -

1. Paper Size

Text pages: 8-1/2 x 11 inches.

Drawings and-graphics::-8-1/2 xy-l inches preferred; however, a larger

size is acceptable provided the finished copy when folded does not exceed

8-1/2 x 11 inches.

2. Paper Stock and Ink

Suitable quality in substance, paper color, and-ink density for handling and for reproduction by microfilming. - , !

... _, .'' . , .*.' _ .

3. Page Margins '

A margin of-no less than 1 inch is to be maintained on the top,-bottom,

and binding side of all pages submitted.

xi
I

*4. Printing

Composition: text pages should be single spaced.

Type face and style: must be suitable for microfilming.

Reproduction: may be mechanically or photographically reproduced. All

pages of the text may be printed on both sides, and images should be

printed head to head.

5. Binding

Pages should be punched for looseleaf ring binding.

6. Page Numbering

Pages should be numbered by section and sequentially within the section.

Do not number the entire report sequentially. (This entire Standard

Format has been numbered sequentially because the individual chapters were

too short for sequential numbering within each section to be meaningful.)

7. Format References

In the application, references to this Standard Format should be by part,

chapter, and section numbers.

Procedures for Updating or Revising Pages

The updating or revising of data and text should be on a replacement page

basis.

The changed or revised portion of each page should be highlighted by a

vertical line in the margin opposite the binding margin for each line changed

or added. All pages submitted to update, revise, or add pages to the report

should show the date of the change. The transmittal letter should include an

index page listing the pages to be inserted and the pages to be removed. When

major changes or additions are made, pages for a revised table of contents

should be provided.

Number of Copies

The applicant should submit the appropriate number of copies of each

required submittal pursuant to § 70.21, "Filing," of 10 CFR Part 70.

Public Disclosure

The NRC has determined that the public disclosure of the details of

physical protection programs is not- in the public interest, and such details

are withheld pursuant to paragraph' 2.790(d) of 10 CFR Part 2, "Rules of Practice

for Domestic Licensing Proceedings." Thus the physical protection section of

each application should be submitted as a separate enclosure. Other proprietary

and classified information should be clearly identified and submitted in separate

enclosures. Each such submittal of proprietary information should be accompanied

by the applicant's detailed reasons and justifications for requesting exemption

from public disclosure as required in paragraph 2.790(b) of 10 CFR Part 2.

xii
PART I

SPECIAL NUCLEAR MATERIAL

OF MODERATE STRATEGIC SIGNIFICANCE

5.59-1
1. USE AND STORAGE AREA AT A FIXED SITE

This chapter provides guidance on meeting the requirements of paragraphs

.73.'67(d)(1) and (d)(2), which are as follows:

(d)(1). .-Use the material only within a controlled access area which is

illuminated sufficiently to allow detection and surveillance of

-unauthorized penetration or activities.*

(d)(2) Store the material-only within a controlled access area such as a

vault-type room or approved security cabinet or their equivalent

which is illuminated sufficiently to allow detection and surveillance

of unauthorized penetration or activities.* .. .

A controlled 'access area (CAM) is defined in paragraph 73.2(z) as "any

temporarily or permanently established area.which is clearly demarcated, "access

to which is controlled and which affords isolation of the material or persons

within it." -'Access control" means measures used to allow-only-specified

personnel,-materials,-and vehicles ingress into and egress from'a given area,

while "isolation" refers to measures taken to deter persons, materials, or

vehicles from entering or leaving a given area through other than'established

access control points.- Thus, a CAM includes'provisions for both isolation and

* 'access control. In some'cases, isolation or access control systems may also

serve the purpose of aiding in the detection of unauthorized penetration or

activities* within the CAM. Therefore, these detection-related considerations

are alluded.to in this chapter.

.,-In the discussion that follows, CAAs intended for use and the possible

storage-ofSNM of.:moderate strategic significance are discussed separately from

those intended solely for the storage-of such SNM. Although the requirements

for these-two different-applications of CAM are in'most-respects similar, it is

recognized that the means used to isolate the material and control access may

differ markedly for use areas compared with-storagegareas and therefore warrant

.separate-discussion. .

1.1 Area Where Material Is Used [73.67(d)(1)'

Intent

This.section discusses CAMs intended primarily for the'use of SNM of

moderate strategic significance. These may be temporarily established to meet

transitory, or intermittent-SNM use requirements or they may be permanently.

established. Permanently established CAAs for use of SMM may also be suitable

for storage. "Use" means that the material is undergoing processing (e.g.,' fuel fabrication, irradiation in a reactor) or utilization of its properties in

conjunction with experimental equipment (e.g., equipment used for research or

educational laboratory experiments). -- Different isolation/access control

measures may be used for periods during which-the area -is occupied versus;

unoccupied. .

Unauthorized activities are those-activities' deemed by the licensee to be

indicative of or contributory to the possible theft of SNM.

5.59-3
Illumination sufficient to allow detection and surveillance of unauthorized penetration or activities within the CAM where the material is used need

not require the use of high-intensity lighting throughout the- CAA. What is

intended is the use of normal lighting sufficiently uniform throughout the CAM

to ensure that-material or unauthorized personnel cannot be- secreted in a

darkened area until a time more convenient for the unauthorized removal of the

material. For those facilities where experiments must be conducted in a

darkened room, the lighting requirement is exempted for as long as is needed

provided access control is ensured and the material is accounted for at the end

of the experiment.

1. Temporarily Established CAAs

Temporarily established CAMs for the use of SNM need not have permanent

barriers at their boundaries. Isolation of the material and persons using the

material may be provided by office partitions, cordons, or other devices used

to warn passersby of the restricted nature of the area. Access control can be

effected through surveillance or supervision of-the area by those who are using

the SNM and who are responsible for the material. However, the provision of

access control through personal supervision is'suitable only for those

situations in which the size of the CAM and the number of persons to be

admitted are sufficiently small and in which the CAM is suitably configured to

make such procedures practical.

When material located in a temporarily established CAM is to be left

unattended (i.e., because it is impractical to replace the material in a CAM

designed for long-term storage of the material), access control and improved

isolation for the temporary CAM may be provided to substitute for the

discontinued personal supervision over the material.- (In addition, provisions

must be made for satisfying the monitoring requirement of'paragraph

73.67(d)(3).) Improved isolation can sometimes be provided by increasing the

penetration resistance of existing barriers (e.g., locking doors and windows,

placing the material in a locked drawer or supply room). Access control may be

provided by any of the measures suggested below for permanently established

CAAs left unattended; some of these suggestions may be more suitable for

temporary use than others. If no suitable barrier is available to provide

isolation, material left unattended may be protected by a motion alarm covering the area immediately surrounding the material. This improved detection

capability would be considered an acceptable substitute for the rudimentary isolation provided by temporary barricades, cordons, signs, and other less

substantial means of isolation that may be used on a temporary basis.

2. Permanently Established CAMs

Permanently established CAMs for the use and temporary storage of SNM of

moderate strategic significance would most likely provide isolation for the SNM

through the use of permanent barriers. These could consist of fences; gates or

freestanding walls for exterior areas; or exterior or interior building walls,

locked doors, windows, bars, grillwork; or other barriers for interior areas.

Such barriers are not required to meet the more stringent criteria for physical

barriers used for the protection of formula quantities of strategic special

5.59-4

<2d
nuclear material (SSNM).* However, good security management practice would

dictate that the barrier be substantial enough to.deter casual passersby from

unauthorized penetration. If the barriers are also designed to aid in detection, the criteria for penetration resistance or tamper indication in accordance

with the specific monitoring procedures to be used-also apply.

Access control for permanently established CMs during periods they are

occupied can be provided by personal supervision. However, when these areas

are unoccupied or when the size,. configuration, or numbers of. persons intended

to be admitted to the area make personal supervision -impractical, other means

of access control may be called for. Some of the additional access control

measures that.may be-used for permanently established CAAs* in these-situations are:

Stationing a watchman at CAA access control points,

Limiting distribution of keys, keycards, or ;ombinations to doors and

gates,

Using a coded badging-system to identify authorized personnel at CAA

access control points,

Controlling locked CM doors and gates by use of remote surveillance

(e.g., intercom or CCTV) by personnel stationed at a centrally located access control facility, or

. Controlling access of personnel, material, and vehicles into CAAs by

,other means. .- - - - -

It is expected that large facilities will have.more complex isolation'and

access control systems than small facilities in order to achieve comparable

levels of protection. . .

Content

(This section needs to be completed only for CMs designated exclusively for use'of'SNM. CAAs 'designated for both use and storage. are to be *Jezrribed

as recommended in Section 1.2. of Part I of this Standard Format.) - -

- Describe the CAAs designated exclusively for useof the material. Indicate under'what conditionsCAAs -will be established on a temporary basis. -For

each CAAidentified, provide the following information . l

. 1. A description of the area andcits features relative to other tacility ;features,'showing the normal routes of. irgress to each CM (including a scale

diagram). - . -

"Strategic special nuclear material" means uranium-235 (contained in uranium

enriched to 20 percent or more in the U-235 isotope), uranium-233, or plutonium.

Access control requirements for CAAs for the use or storage of SNM of 'moderate

strategic significance are discussed in Chapter 3, "Access Control at a Fixed

Site," of Part I of this guide.

5.59-5
2. Descriptions of physical barriers, cordons, walls, partitions, or

other means of providing isolation of the CAM and channelling entry through

established access control points into the CAA.

3. The means and criteria used for controlling access into the CM at

established access control points.

4. If more than one CAM is designated, the types of material normally

used in each CAA; for temporarily established CAAs, the types of activities

normally performed within each area.

5. The lighting level and uniformity of lighting provided to allow

detection and surveillance of unauthorized penetration or activities within the

CAM or in the immediate vicinity of the CAA.

1.2 Area Where Material Is Stored [73.67(d)(2))

Intent

CAAs normally used only for storage of SNM are described in this section.

Such CAMs may be very limited in size since they need only contain the material

itself, and access to them is expected to be infrequent. Where small amounts

of material are involved, the CAA may consist of relatively small containers

such as security cabinets, safes, and locked closets or supply rooms.

The same basic isolation and access control* capabilities that apply to

CAAs for the use of SNM-of moderate strategic significance also apply to

CAAs for the storage of this material. However, the CAM for storage is, in

addition, specifically required to be equivalent to either an approved security

cabinet or a vault-type room. This additional requirement embodies a tradeoff

decision to be made by the licensee between the capability for immediate

detection of penetration attempts into a vault-type room and the'improved

penetration resistance likely to be provided by an approved security cabinet.

A vault-type room is defined in paragraph 73.2(o) of 10 CFR Part 73 as "a

room with one or'more doors', all capable of being locked', protected by an

intrusion alarn which creates'an alarm upon the entry'of a person anywhere into

the room and upon exit from the room or upon movement of an individual within

the room." The vault-type room can be a locked laboratory, supply room,

closet, or other room equipped with a tamper-resistant motion detector alarm

system. The motion detector would simultaneously satisfy the monitoring

requirement of paragraph 73.67(d)(3) addressed in the next chapter. The

expression "equivalent to a vault-type room" means that a piece of equipment

(such as a fission'chamber, reactor core, or storage rack), even though it does

not resemble a "room," may meet the requirement for a storage-type CAA if there

is a means of isolation (e.g., a locked grill, inaccessibility beneath water as

in a storage pool, intricate and time-consuming procedures required for

removal) and it is protected with a tamper-resistant motion detection system.

Access control requirements for CAAs for the use or storage of SNM of moderate

strategic significance are discussed in Chapter 3 of Part I of this guide.

5.59-6

'<-I
Note that, in some cases, material located in such pieces of equipment may be

-considered in use rather than in storage. In this case, whether the material

is being personally supervised or not," the requirement for use of a motion

detection system to satisfy the'monitoring requirement is removed, and other

monitoring devices or procedures may be used instead.

When a container equivalent to an approved security cabinet is chosen to

satisfy paragraph'73.67(d)(2), the equivalency is based on the penetration

resistance of 'the container and the difficulty associated with manipulating the

lock. The basic premise'used to determine whether such a container may be,

monitorediby procedures-or devices other than a-motion detector is whether it

is unreasonable to expect that an external adversary could penetrate the

container in a reasonable amount of time without leaving an indication of the

penetration. The amount of time that would be required for penetration without

such an indication may be used to govern the frequency of patrols when

monitoring procedures are used.' An approved security cabinet or its-equivalent

is one whose design has-been certified by the General Services Administration

or other nationally recognized standards organization (e.g., ANSI) to afford

protection 'against surreptitious entry and lock'manipulation'equivalent to that

provided by'a Class-6 GSA rating or better.

Isolation for CAAs intended only for storage usually will be provided by -- the penetration-resistant features of the perimeter barriers'or container walls

of the CAM. 'The 'level 'of penetration resistance of such barriers is 'not'

required to meet the more stringent criteria for physical'barriers used'to'

protect formula quantities of SSNM. Rather, the physical barriers or 'container

walls 'of the CAM are intended only to deter penetration by unauthorized persons

and to aid detection by providing an indication of forced penetration.

"In determining the level of isolation/access control required for such

CAAs, the time required for extricating'SNM from its'storage location'in , relationship to the licensee's time of detection and assessment may be taken

into account (e.g., spent fuel assemblies'stored in spent fuel storage pools or

7 fuel residing in nonpower reactor cores).,

Examples of typical CAAs where special nuclear material of moderate

strategic significance may be stored are:

1. Vault. A structure that satisfies the definition for a vault as

stated in paragraph 73.2(n) would provide more than adequate isolation

protection.

2. Approved Security Cabinets. Those cabinets that are designed to

provide delay and resistance against surreptitious entry and lock manipulation.

3. Reactor. Reactors that are so designed that removal of material is

difficult.

4. Vault-Type Room. Some typical vault-type rooms where materials are

stored and protected with a motion detector are storage pools, a room

containing in-process storage racks, and laboratories where material is left

unattended. In all cases, movement in the near vicinity of the material or of

the material itself generates an alarm signal.

5.59-7
S. Locked Laboratories, Supply Rooms. Theserareas must be sufficiently

penetration resistant-to afford a means of isolation/access control and permit

the proper functioning of the system for monitoring the storage area as

required by paragraph 73.67(d)(3) of the rule and as described in Chapter 2,

"Detection Devices and Procedures at a Fixed Site," of Part I of this guide.

The illumination level required for the CAM should be sufficiently uniform

and bright (a) in the case of a security cabinet, to detect penetration of or

tampering with the CAM containment or (b) in the case of a vault-type room or

its equivalent, to detect unauthorized penetration of or activities within the

CAM.

Content

Describe the CAMs where the material will be stored and, in some cases,

used. For each such CAM, provide the following information:

1. A description of the area and its features relative to other facility

features, showing the access control points for the CAM (including a scale

diagram).

2. The-isolation system such as physical barriers, container walls, or

other features that demarcate the perimeter of the CAM and channel entry only

through established access control points.

3. The means and criteria used for controlling access to the CAM at

established access control points.

4. If more than one CAM is designated for the storage of the material,

the types of material normally to be stored in each area.

5. The lighting level and uniformity of lighting provided to allow

detection and surveillance of unauthorized penetration or activities within the

CAM or in the immediate vicinity of the CAA (where security cabinets are used).

5.59-8
.2. DETECTION DEVICES AND PROCEDURES AT A FIXED SITE

This chapter provides guidance for meeting the'requirement of paragraph

73.67(d)(3), which is as follows:

*(d)(3) Monitor with an intrusion alarm or other device or procedures the

controlled access areas to detect unauthorized penetrations or

activities.

The purpose of this monitoring activity as stated in paragraph 73.67(a)(2)

is to provide "early detection and assessment of unauthorized access or activities by an external adversary within the controlled access area..."'and "early

detection of removal of special nuclear. material, by an external adversary from

a controlled access area.". This should be done to achieve the objective of

minimizing the possibilities for unauthorized removal of special nuclear

. material "consistent with the consequences of such actions," as'stated 'in

paragraph 73.67(a) of the effective.rule. Thus, the earliness.of detection

afforded by the licensee's physical protection system may vary depending on the

nature and quantity of the material. susceptible .to unauthorized'removal.

Further discussion of the earliness of detection required for possible thefts

of different types and quantities of special nuclear material of moderate

strategic significance is provided in Section 2.1 of Part%.I of this guide.

2.1 Earliness of Detection [73.67(a)] -

Determination of early detection of unauthorized.access, activities, or

removal of'SNM from CAAs will'be based on an assessment of the magnitude of the

consequences associated-with possible misuse ofthe type and quantity of

material that could be removed in a given theft attempt. For SNM of moderate

strategic significance, two.distinct cases are considered:.

1.' Theft of-strategic special-nuclear material (SSNM).* -

2. .Theft of low-enriched uranium (LEU).**

Thefts from a single facility of SSNM in quantities of moderate strategic

significance are limited by definition to quantities that could not be used to

construct a nuclear~explosive device;-thefts of similar quantities of material

from several different facilities could, however, lead to the accumulation of

an aggregate quantity that would permit such illicityuse. (The consequences of

a single theft'of such'material would have minimal impact on the public health

and safety.) Therefore,.,detection of such-thefts is required sufficiently

early to, ensure-early notification to the NRC so that the NRC in turn can

notify other licensees of.the need to'-implement appropriate-responses to

prevent the' accumulation' by a'single adversary of a formula quantity of

material through multiple thefts from.different facilities. The NRC believes

Athat there are detection systems'and.procedures that would allow the licensee

to detect a theft of.SSNM within approximately 2.hours..

: In. this case,, strategic special-nuclear material'means all 'SNM of moderate

strategic significance other than low-enriched uranium.

Low-enriched uranium is uranium enriched above natural percentages but less

than 20 percent in the U-235 isotope.

5.59-9
The criteria for early'detection of thefts of LEU enriched to more than

10 percent, but less than 20 percent, in the isotope U-235 are the same as for

thefts of LEU in-quantities comprising SNM of low strategic significance. The

reader should refer to Part II of this guide for information regarding

appropriate monitoring procedures to achieve early detection for such material.

2.2 Detection Throuqh Monitoring Controlled Access

Areas [73.67(d)(3)]

Intent

Licensees possessing quantities of SNM of moderate strategic significance

can provide a monitoring system capable of detecting a theft of material within

approximately 2 hours of the actual removal of the material. Either detection

devices or security procedures may be used to help detect unauthorized activities

or penetration of CAAs.- Most of the sites at which this'material is stored or

used are likely to be small research-oriented facilities such as nonpower

reactors and educational institutions. At such facilities there are likely to

be relatively small quantities of SNM available, and authorized access to it

may be infrequent. -

Detection devices such as interior motion detector systems, balanced

magnetic switches, etc., could be used to monitor such relatively small CAAs

where the number of established access control points is very limited.

Security procedures may also be used to protect against thefts of quantities of SHM of moderate strategic significance because watchman patrols

inherently allow unguarded intervals during which an external adversary could

conceivably obtain access to-the CMA, remove a small quantity of material, and

leave before being discovered.' The inspection procedure may (1) be sufficiently frequent to ensure that an external adversary could not complete a

successful theft'of the material during the unguarded intervals, (2) provide

for an indication of the CAA barrier having been tampered with (as with a seal

or by observation of damage done to the barrier), or (3) provide for an item

count inventory or other accounting for the material as part of the inspection.

Some examples are given below:

After having left material unattended in an experimental setup in a

classroom used 'as a temporary CAA, the instructor returns after

several hours and verifies that the material is still in place.

A small quantity of material is stored in a security cabinet for

several months'at a time without authorized access. 'Every 2 hours a

watchman comes around to inspect the cabinet to ensure that the lock

Is secure and the cabinet has not been tampered with. 'If a lock is

used that is not designed to be resistant to surreptitious and forced

entry, the watchman also inspects a tamper-indicating seal to verify

that the container has not been tampered with.

During periods when a nonpower reactor is left unattended, a watchman

comes around every 2 hours and checks that none of the fuel elements

5.59-10
have been removed by unauthorized persons. If a person working

during off-hours'in the vicinity of the nonpower reactor is

encountered, appropriate identification is 'requested.

While a nonpower reactor is being used, the supervisor ensures that

-he/she or his/her designee-is not away from the reactor for a period

-longer than 30 minutes or alperiod he/she estimates would be

minimally required for a person to obtain fuel elements 'in the core

that are not self-protecting by virtue of their having been

sufficiently irradiated.

During periods when excess fuel is:left unattended in a storage

closet at a nonpower reactor facility, an alarm'system is activated

that would annunciate if anybody tried to penetrate the closet

without proper authorization. :

During periods when the material is being used under the direct supervi-

--sion-of authorized personnel,-the monitoring requirementmay be satisfied by

virtue of the continuous surveillance exercised by-such personnel.

Content

For each of the CAAs designated for the protection of SNM of moderate

strategic significance, provide the following information regarding the measures taken to fulfill the monitoring requirement of paragraph 73.67(d)(3):

1. Describe any devices employed by specifying the type of device, its

installed location, the type and location of annunciation, the intended area of

coverage, and the tamper-resistant features. 'Refer, if desired, to existing

filed security plans, regulatory guides, or NUREG documents.

2. Describe any procedures employed to monitor the CAA or portions

thereof, including the categories of persons who will execute the procedures,

the frequency of inspections or rounds of patrol, the basis for the determination of such frequencies, and the occasions when the procedures are intended to

be implemented.- Also indicate the features of the CAM (e.g., barriers, locks,

seals) that will affect the way the procedures are used. - -

- . . . ..

3. Explain how the combination of procedures-and devices used to monitor

each CAA meets the criteria for early detection of theft by an external

adversary relative to the type of material found in -each CAM.

* ' I -$, . . ' - . .

5. 59-11
- 3. ACCESS CONTROL AT A FIXED SITE

This chapter provides guidance on meeting the requirements of paragraphs

73.67(d)(4), (d)(5), (d)(6), (d)(7), and (d)(10), which are as follows:

(d)(4) Conduct screening prior to granting an individual unescorted access

to the controlled access area where the material is used or stored,

in order to obtain information on which to base a decision to permit

such access,

(d)(5) Develop and maintain a controlled badging and lock system to identify

and limit access to the controlled access areas to authorized

individuals,

(d)(6) Limit access to the controlled access areas to authorized or escorted

individuals who require such access in order to perform their duties,

(d)(7) Assure that all, visitors to the controlled access areas are under the

constant escort of an individual who has been authorized access to

the area,

(d)(10) Search on a random basis vehicles and packages leaving the controlled

access areas.

An access control system is one that controls access to or egress from a

CAA through normal routes for Dersonnel, materials, or vehicles.

3.1 Preauthorization Screening [73.67(d)(4)]

Intent

The intent of the requirement for preauthorization screening is to ensure

that the licensee will have-sufficient knowledge of an individual to determine

his/her reliability and need for access prior to granting him/her authorized

access to the CAM where the material is used or stored. The selection of

procedures for conducting this examination and the criteria employed to make

these judgments are the responsibility of the licensee and, of course, should

be consistent with all local, State, and Federal laws and regulations regarding

the protection of the privacy'and other rights of the individual. The

screening process may be'conducted in the same manner as other investigations

customarily conducted by potential employers for'similarly sensitive positions.

There is no requirement for the licensee to arrange for an NRC clearance or

similar clearance from any other government organization. Examples of

procedures and criteria that may be employed in the screening process include

holding or having recently held a government-sanctioned clearance; examination

of past employment or educational records (to determine any unsatisfactory

employment or school actions or incidents that would indicate any unreliability

or previous breaches of trust between the individual and his/her employer);

endorsements or references from previous employers, teachers, or colleagues

that would support a decision for granting access or that would attest to the

trustworthiness and reliability of the individual; and consideration of the

individual's present employment record indicating demonstrated trustworthiness

5.59-12
and reliability over an extended period of employment with the licensee. (This

may be considered in the nature of "grandfathering.")

Content -

Describe the procedures and criteria that will be used for obtaining sufficient information prior to making a decision on granting unescorted access

authorization to an individual to CAAs where the material is used or stored.

Identify the types of individuals who will be screened (e.g., process engineers, supervisory personnel, professors, instructors, graduate students)

and who will perform the screening process.'

3.2 Badging System [73.67(d)(5)]

-.Intent -

The purpose of the badging system is to facilitate the identification of

authorized individuals and the control of access to or within the CAA where the

material is used or stored. Information on the badge should be such that it is

possible to clearly distinguish personnel authorized for access to the CAAsfrom those requiring an escort. Information on the badge should also uniquely

identify the individual possessing the badge. This personalized information

can be obtained through the use of photographs, personal-vital statistics,

signatures, or any means the licensee may wish to use that will uniquely

identify the individual.

Content.

Describe the badging system used to facilitate control of access to the

CAAs. This description should include: ;

1. The size, shape, color, material, and construction of badges.'

2. The distinguishing features of the badge that'identify authorized

individuals from escorted individuals.

3. How the badges will be used for controlling access.' (For example,

will all individuals be checked'prior to entering the CAAs,' will periodic' ' checks be made of individuals within a CAM to determineif they'are authorized'

*or under escort, or will the badge itself permit authorizedentrance, e.g., a

card key.) -

4. The system used for issuing, controlling, and accounting for the

badges. - - ' - '

3.3 Lock System [73.67(d)(5))

Intent -

Locks used to control access to CMWs should be resistant to manipulation

or picking and'should not be mastered. Exiaples of typical, lock systems that

fit this description are three-iosition dial-type' combination locks, six-pin. key locks, and card-key lock systems. The procedures for assigning keys and

combinations to individuals is an integral part of the lock system and should

5.59-13
be designed to ensure that only authorized personnel- have access to such items.

Locks and combinations should be changed when information is obtained that the

lock system may have been compromised. Further information may be obtained in

Regulatory Guide 5.12, "General Use of Locks in the Protection and Control of

Facilities and Special Nuclear Materials."

Content

Describe the locking system used to control access to the CAMs where

material is used and stored. This description should include locations of all

locks included in the system by type of lock, the pick-resistant and

manipulation-resistant characteristics of each lock type used, personnel

responsible for issuing keys or combinations and changing combinations or

locks, criteria for changing combinations or locks, personnel authorized to be

given keys or combinations, and descriptions of types of locks used (references

may be made to Regulatory Guide 5.12 for this purpose).

3.4 Personnel Entry Control System [73.67(d)(6)]

Intent

The success of other access control system components such as preauthorization screening, badging, and lock control is dependent upon effective control

of personnel access into the CAM. Physical access may be controlled in a

number of different ways depending on the actual configuration of the CAM and

other site-specific factors. Some examples of these alternatives are:

1. Control by Authorized Person. If the area to be controlled is

sufficiently small and free of obstructions, an authorized person performing

other activities in a CM may effect physical. access control by monitoring

entry of unauthorized persons into the area. A sign posted at the entrance

would help deter casual passersby.. An example of this approach would be when

laboratory instructors are conducting classes and, because they are familiar

with each of their students, they could easily recognize unauthorized persons

not belonging in the classes.

2. Card-Key. Cipher, Combination, or Key-Lock Control System. A more

sophisticated hardware-oriented system involves the use-of a card-key, cipher,

combination, or key-lock system. Physical access control in this case consists

of the use of physical barriers to deter unauthorized persons. A limited

number of entrances that are controlled by authorized personnel using a card

key, cipher, combination, or key are provided. This system may be more useful

when larger numbers of authorized personnel who would not necessarily be

familiar with one another need to share the use of the CAA.

3. Control by Security Organization. If security organization personnel

are available, physical access control may be accomplished by stationing a

person at the entrance to the CAM to check identification and allow only

authorized persons into the CM.. This alternative may be unjustifiably

expensive unless the security organization member's salary can be justified on

other grounds as well. A variation of this system requires persons seeking

entrance to the CAA to obtain a key from a properly designated person or

security organization for each use.

5.59-14

KJ-_
Content

Describe the system for limiting physical access to each'CAM identified in

Sections 1.1 and 1.2 of Part I to authorized personnel or those escorted by

authorized personnel. Include in this description the names or titles of

individuals granting access, authorizations, the criteria to be used in granting

authorizations, and the procedures used to ensure that-only authorized or

properly escorted persons are allowed access'to the CAA. Reference may be made

to Sections 3.1, 3.2., 3.3, 3.5, and 3.6 of this'chapter'as they apply to this

section for the description of locks, barriers, or other hardware thatare used

to control access.

3.5 Escort System [73.67(d)(7)3-

Intent

- The'reqirement that an escort system be established is in recognition of

the fact'that the licensee may wish to-allow access to-certain persons or

classes of persons on a temporary. or infrequent basis or on 'short''notice , thus

making the routine process for granting access authorizations impractical or

inexpedient. Typical arrangements :for escorted access may include escorts for

maintenance or repair personnel, laboratory classes, public'tours, guests, and

visitors as required. l - -

Content

Describe the system that will be used to escort individuals in the CAAs.

In its security plan, the licensee should ensure that only properly authorized

individuals will be allowed to escort individuals. This description should

include:

1. Criteria to be used for granting escorted access,

2. Criteria to be used for escorting others,

3. Procedures for escorting individuals into CAAs (e.g., students under

the supervision of laboratory instructor, public tours),

4. The number of escorted individuals per escort, and

5. The responsibilities of the escort (e.g., periodic surveillance of

all individuals under escort, accounting for all material prior to leaving the

CAA, remaining in general area during the time unauthorized individuals are

present).

3.6 Search [73.67(d)(10)]

Intent

The primary intent of the search requirement is to deter and possibly

detect attempted thefts of SNM. The search procedures developed by the

licensee should take into consideration the environs where the material is used

5.59-15
or stored, the physical characteristics of the material itself, and the

frequency of accounting for the material. In some cases, this will require

that all vehicles and packages leaving the CAAs be searched in a random manner.

The frequency of random searches should be determined by the ease with which

the material can be stolen and'the length of time it would take to detect a

theft. In'other cases, only packages that equal or exceed the size of the

material being used or stored would have to be searched, taking into

consideration the difficulty with which the material could be broken into

smaller, more easily concealed parts.

Content

Describe the system to be used for randomly searching vehicles or packages

that leave the CAA. Include in the description information as to:

1. The scope of the search. This should identify the criteria that will

be used for searching vehicles and packages (e.g., whether all packages and

vehicles are subject to search or just those packages or vehicles that are

larger than the smallest configuration of material being used or stored).

2. The randomness of the search.' The scheme for selecting the packages

or vehicles to be searched should be identified (e.g., subjecting each package

or vehicle to a search, using a random number generator for determining whether

a candidate package or vehicle is to be searched, searching a minimum

percentage of all packages or vehicles leaving the CM each day).

<-

5.59-16
4. SECURITY ORGANIZATION AT A FIXED SITE

This chapter provides guidance on meeting the requirements of paragraph

73.67(d)(8), which states:

(d)(8) Establish a security organization or modify the current security

organization to consist of at least one watchman* per shift able to

assess and respond to any unauthorized penetrations or activities-in

the controlled access areas.

Intent

The intent of this requirement is to ensure that, in the event'of a

security incident, someone will be available to assess-alarms or otherunauthorized penetrations or activities and,, if warranted, notify the NRC, the

local law enforcement authorities, and the responsible person in licensee

management. Early detection and notification of any missing material will help

facilitate its prompt recovery' In some cases, the licensee may assign

additional duties to members of.the security organization where procedureoriented options are chosen to' satisfy physical protection requirements (e.g.,

periodic patrols and inspections of CAAs for storage of SNM). .Security organization members' are not required to be fully dedicated full-time employees of

the licensee. They may include unarmed campus security. personnel (watchmen),

contract guards, members of the local law-enforcement agency (if sufficiently

close'to the site), etc. No formal or comprehensive training program is

required for security organization personnel. However, the licensee should be

prepared to demonstrate that each security person understands the particular

duties assigned to him/her and is fully qualified and trained to perform them.

Content - - - ;

Describe the security organization that will be responsible for assessing

and responding to security incidents. Indicate the other responsibilities of

the security organization such as:

1. Conducting periodic physical security checks of CAAs,

2. Maintaining liaison with the local law enforcement agency,

3. Notifying the local law enforcement agency of any unauthorized

penetrations or activities in the CAMs, and

4. Notifying licensee management of any unauthorized penetrations or

activities in the CAAs.

A "watchman" is defined in paragraph 73.2(d) of 10 CFR Part 73 as "an individual, not necessarily uniformed or armed with a firearm, who provides protection for a plant and the special nuclear material therein in the course of

performing other duties."

5. 59-17
5. COMMUNICATIONS AT A FIXED SITE

This chapter provides guidance on meeting the requirements of paragraph

73.67(d)(9), which states:

(d)(9) Provide a communication capability between the security organization

and appropriate response force.

Intent

The Intent of this regulation is to ensure that a communication capability

exists between the licensee and the designated response force. It is implied

that, prior to setting up a communication capability, procedures and

responsibilities will have been established between the response force and the

licensee. (See Chapter 6, "Response Procedures at a Fixed Site," of Part I.)

The type of communication system chosen by a licensee should:

1. Provide for full duplex voice communication capability,

2. Be easily accessible to the licensee's security organization, and

3. Be reliable and available for immediate use at any time.

Some communication systems that would provide these capabilities include a

dedicated telephone system, a nondedicated public telephone system, radio, or

any combination thereof.

Content

Describe the communication system that is used between the security

organization and the appropriate response force. This description should

include information on:

1. Type of communication system,

2. Location of voice terminals in relationship to CAAs,

3. Availability of communication system on a 24-hour basis, and

4. Reliability of communication system.

5.59-18
6. RESPONSE PROCEDURES AT A FIXED SITE

This chapter provides guidance on'meeting the requirements of paragraph

73.67(d)(11), which states:

(d)(n) Establish and maintain response'procedures -for -dealing with threats

'of thefts or thefts of such materials.

Intent

-The intent of this regulation is to help, the licensee to identify those

security incidents that could result -in the loss of'SNM of moderate strategic significance and to develop response procedures to prevent or reduce the

likelihood of such a loss. Some types of incidents that, should be considered

and-for which response procedures should be developed are:

1. -Situations that could possibly lead to theft of SNM (e.g., civil

strife),

2. Discovery that the security system has been breached,.and

3. Discovery that some SNM is missing.

Content

Identify those events for which' response procedures will be developed. Also, describe the type of response to belaccomplished,for each event identi-.

fied and the duties and responsibilities of -the security organization and

management involved in the response. Ensure that the NRC will be notified

immediately in the event of theft or attempted theft of the material. Describe

what local law enforcement assistance is available, their response capabil-:

ities, and any agreements made with them to respond in, the'case of'theft of the

material. '

5.59-19
7. MATERIAL TRANSPORTATION REQUIREMENTS

This chapter provides guidance on meeting the requirements of paragraph

73.67(e)(1), which are as follows:

(e)(1) Each licensee who transports, exports, or delivers to a carrier for

transport special nuclear material of moderate strategic significance

shall:

(i) Provide advance notification to the receiver of any planned

shipments specifying the mode of transport, estimated time of

arrival, location of the nuclear material transfer point, name

of carrier and transport identification,

(ii) Receive confirmation from the receiver prior to the commencement

of the planned shipment that the receiver will be ready to

accept the shipment at the planned time and location and

acknowledges the specified mode of transport,

(iii) Check the integrity of the container and locks or seals prior to

shipment, and

(iv) Arrange for the in-transit physical protection of the material

in accordance with the requirements of paragraph 73.67(e)(3) of

this part unless the receiver is a licensee and has agreed in

writing to arrange for the in-transit physical protection.

7.1 Advance Notification to Receiver [73.67(e)(1)(i)]

Intent

The intent of this paragraph is to require the shipper to preplan the

transportation of material and inform the receiver of his plans prior to

shipment. This is the first of the several transportation requirements that

will allow the receiver to take delivery of the material as planned or to help

ensure traceability of any missing material.

Content

The licensee should ensure in his security plan that, pridr to each

shipment of material, the receiver will be notified of the impending shipment

and provided the following types of information:

1. Mode of transport (e.g., truck, plane, train, or ship),

2. Estimated time of arrival,

3. Location where custody of the material will be transferred to the

receiver,

4. Name of carrier, and

5. 59-20
5. Transport identification (e.g., truck, train, or flight number; ship

name).

7.2 Receiver Confirmation r73.67(e)(I)(ii)]

Intent

The intent of this requirement is that, prior to shipment, the shipper

will be assured that the receiver is ready -to accept the shipment at'the

planned time and location and has acknowledged the mode of transport.

Content

Describe what'procedures will be used to ensure that shipment of material

does not take place until the receiver acknowledges the planned shipment and

mode of transport and.states that he will be ready to accept the shipment at

the planned time and location.

7.3 Inspection [73.67(e)(1)(iii)]

Intent

The intent of this paragraph is to provide a means for confirming the

integrity of a shipment through inspections just prior to commencement and

during the shipment by ensuring that the'material containers and any associated

locks or seals are intact at the time the shipment commences. '

Content

Describe the procedures to be used to ensure that the integrity of the

shipment containers and associated locks or seals is checked-just prior to

shipment.'

7.4 Responsibility for In-Transit Physical Protection

-- -- 7367(e)(1)(iv)]

Intent

The intent of this paragraph is to make clear that the licensee.shipping

the material is responsible for arranging for the'physical protection of the

material in transit if the receiver is not a licensee. If both the shipper and'

receiver are licensees, the shipper may allow the receiver to accept this

responsibility wholly or in part, provided there is appropriate documentation

specifying their respective responsibilities. Where no such documentation

exists,.both-the shipper and receiver are held jointly responsible. (See

Section 8.3 of Part I of this guide.)' '

Content'

In its security'plan,- the'shipper sho'uld'either acknowledge responsibility

for the in-transit physical protection of SNHM of moderate strategic significance or ensure that a written agreement from the receiver licensee has been

received in which the receiver accepts either full responsibility or shared

responsibility for the in-transit physical protection of this material in

accordance with paragraph 73.67(e)(3) of lO CFR Part 73. -

5.59-21
8. RECEIVER REQUIREMENTS--TRANSPORTATION

This chapter provides guidance on meeting the requirements of paragraph

73.67(e)(2), which are as: follows:

(e)(2) Each licensee who receives special nuclear material of moderate

strategic significance shall:

(i) Check the integrity of the containers and seals upon receipt of

the shipment,

(ii) Notify the shipper of receipt of the material as required in

Section 70.54 of Part 70 of this chapter, and

(iii) Arrange for the in-transit physical protection of the material

in-accordance with the requirements of paragraph 73.67(e)(3) of

this part unless the shipper is a licensee and has agreed in

writing to arrange for the in-transit physical protection.

8.1 Inspection t73.67(e)(2)(i)1

Intent

This requirement is intended to determine whether the material's container

has been compromised en route and whether any material has been removed so that

immediate recovery procedures may be initiated if required.

Content

Describe the procedures to be used to ensure that the integrity of the

containers and seals will be checked upon receipt of the shipment of material.

8.2 Notification to Shipper [73.67(e)(2)(ii)]

Intent

This requirement is intended to provide the shipper with knowledge of when

responsibility for the shipment has been assumed by the receiver and when the

shipper's responsibility is ended.

Content

Ensure that the receiver will send a completed copy of standard Form

NRC-741, "Nuclear Material Transaction Report," to the shipper within 10 days

of receiving a shipment of material as required in S 70.54 of 10 CFR Part 70..

8.3 Responsibility for In-Transit Physical Protection

Intent

The intent of this paragraph is to make clear that the licensee receiving

the material is responsible for arranging for the physical protection of the

5.59-22

'\'j
material in transit if the shipper is not a licensee. If both the shipper and

receiver are licensees, the receiver may allow the shipper to accept this

responsibility either wholly orin part provided.there is appropriate

documentation specifying their respective responsibilities. Where no such

documentation exists, both the shipper and receiver-are held jointly

,responsible. (See Section 7.5 of Part I of this guide.)

Content

In its security plan, the receiver should either acknowledge responsibility for ,the:in-transit physical protection of SHM of moderate-strategic signif-

.icance or ensure thatc awritten agreement from the shipper, has been received in

-which the shipper accepts either full responsibility or shared responsibility

for the in-transit.physical protection of this material in accordance with

paragraph 73.67(e)(3) of 10 CFR Part 73.

. ..

, . . . .

5.59-23
9. IN-TRANSIT PHYSICAL PROTECTION REQUIREMENTS

This chapter provides guidance on meeting the requirements of paragraph

73.67(e)(3), which are as follows:

(e)(3) Each licensee who arranges for the in-transit physical protection of

special nuclear material of moderate strategic significance or who

takes delivery of this material free on board (f.o.b.) the point at

which it is delivered to a carrier for transport shall:

(i) Arrange for telephone or radio communications between the

transport and the licensee or its designee: (A) to periodically

confirm the status of the shipment, (B) for notification of any

delays in the scheduled shipment, and (C) to request appropriate

local law enforcement agency response in the event of an

emergency.

(ii) Minimize the time that the material is in transit by reducing

the number and duration of nuclear material transfers and by

routing the material in the most safe and direct manner,

(iii) Conduct screening of all licensee employees involved in the

transportation of the material in order to obtain information on

which to base a decision to permit them control over the

material,

(iv) Establish and maintain response procedures for dealing with

threats of thefts or thefts of such material,

(v) Make arrangements to be notified immediately of the arrival of

the shipment at its destination, or of any such shipment that is

lost or unaccounted for after the estimated time of arrival at

its destination,

(vi) Initiate immediately a trace investigation of any shipment that

is determined to be lost or unaccounted for after a reasonable

time beyond the estimated arrival time.

(vii) Notify immediately the Director of the appropriate Nuclear

Regulatory Commission Regional Office listed in Appendix A to

this part of the action being taken to trace the shipment.

9.1 Communications [73.67(e)(3)(i)]

Intent

The primary intent of this requirement is to ensure that procedures are

established or facilities made available to permit communications to take place

between the transport and the licensee or its designee for the purposes of

(l) periodically confirming the status of the shipment while it is in progress,

(2) notifying the licensee of delays along the shipment route that may affect

the shipment's established schedule, and (3) requesting assistance from local

5.59-24

K\-/
law enforcement authorities (LLEAs). These communications need not be direct

between the licensee and the transport.

1. Confirmation of Shipment Status

Communications for confirming thestatus of a shipment should be timed'to

reflect the expected changes in shipment status indicated by the shipment's

scheduled itinerary. For general motor freight (involving less-than-truckload

quantities), direct communications between the transport and the licensee may

be impractical. This requirement may then be satisfied by periodically

telephoning the carrier to obtain updated reports on.the..shipment's,progress

based on calls the carrier receives from drivers on a. regular basis. ;The

frequency with which the licensee should check the status of a.shipment should;

normally be at least every 10 to 14 hours. This corresponds to the'frequency

with which the drivers normally would call in._: Similar, indirectcommunications

may be used to satisfy this requirement in the case-of air shipments, but they

should be more frequent to reflect the shorter air travel times between

terminals. When direct radio or telephone communications are practical between'

the licensee and the transport, the frequency of calls-should normally be

approximately every 12 hours.

2. Notification of Shipment Delays

Notifications tothe licensee should be made by -the carrier or transport

personnel whenever, unforeseen conditions arise thatthreaten to delay the ,' - shipment significantly beyond the estimated arrival time~at the next point on-,

the shipment's itinerary. This notification may be made by'conventional

telephone lines or by radio.. Installation of radiotelephone lines or,-

comparable equipment is'not required. These communicationsneed not be direct

between the'transport and the licensee or its designee. The-transport

personnel may relay the information to the licensee through the carrier's

offices.

3. Requests for LLEA Assistance

Procedures or facilities are also required to facilitate the transport

personnel in obtaining assistance from LLEAs. Use of radiotelephone or,-

citizens' band radio equipment for this purpose is acceptable but not required.

The transpo'rt''personnel may alternatively use conventional,'telephone',lines. - Requests for assistance should.be in accordance with the response procedures

developed to'comply with paragraph'73.67(e)(3)(iv). s proced'res

Content

Describe the communications'facilities and procedures to be used to ensure.

that communications are established to achieve the three-purposes enumerated in

paragraph 73.67(e)(3)Ci). The'names and telephonenumbers'of responsible

individuals should be given where appropriate. Refer to established response

procedures (Section 9.4 of Part I of this'guide)-as necessary.,

5.59-25
9.2 Minimum Transit Times [73.67(e)(3)(ii)]

Intent

This requirement is intended to have the shipper or receiver. make a

reasonable effort to ship the material by the fastest and most direct method

possible. It is not intended to require use of dedicated transports or other

expensive modes of travel.

Content -

Describe the procedures and considerations that apply in the transportation planning process to ensure that a determined'effort will be made to

minimize transit times.

9.3 Preauthorization Screening [73.67(e)(3)(iii)]

Intent

The intent of the requirement for preauthorization screening is to ensure

that the licensee will have sufficient knowledge of an individual to determine

his/her reliability and need for access prior to granting him/her authorized

access to the material in transit. The selection of procedures for conducting

this examination and the criteria employed to make judgments are the

responsibility of the licensee and; of course, should be consistent with all

local, State, and Federal laws' and regulations regarding the protection of the

privacy and other rights of the individual. The screening process may be

conducted in the same manner as are other investigations customarily conducted

by potential employers for similarly sensitive positions. There is no

requirement for the licensee to arrange for an NRC clearance or similar

clearance from any other government organization. Examples 'of procedures and

criteria that may be employed in the screening process include holding or

having recently held a government-sanctioned clearance; examination of past

employment records (to determine any unsatisfactory employment or incidents

that would indicate any unreliability or previous breaches of trust between the

individual and his/her employer); endorsements or references from previous

employers or colleagues that would support A decision for granting access or

that would attest to the trustworthiness and reliability of the individual; and

consideration of the individual's present employment record indicating demonstrated trustworthiness and reliability over an extended period of employment

with the licensee. (This may be considered in the nature of "grandfathering.")

Content

Describe the procedures that will be used for obtaining sufficient

information prior to making a decision'on granting unescorted access

authorization to those licensee'employees who will be directly involved in the

transportation or in the planning and movement control of the material.

Identify by title or name those employees who will be screened and those who

will perform the screening process.

5.59-26
9.4 Response Procedures [73.67(e)(3)(iv)]

Intent

The intent of this regulation is to help the licensee to identify those

transportation incidents for which notification might be expected and thatmight affect the security of the .SNM in transit and to plan response procedures

for. such situations. -For example, if the shipper is informed by the carrier

that adverse weather -conditions have temporarily prevented further progress of

the shipment, the licensee should inform the receiver of a'new estimated time

of arrival.

Content $ t1

; Identify those~events for which'response procedures'will be developed.

Also, describe types of~response to be accomplished for each event'identified

and the duties and responsibilities of members 'of the-security organization and

management for dealing with the response. Ensure that the NRC will be notified

immediately in the event of theft or attempted theft of the material.

9.5 Notification (73.67(e)(3)(v)3

Intent

The intent of this requirement is to ensure that the licensee responsible

for the physical protection of SNM in transit will have a firm basis for

deciding whether or not to initiate response procedures in the event a shipment

becomes overdue or is lost.

Content

Describe the arrangements and procedures that will be used for notifying

the licensee who arranges for the physical protection of material in transit of

the arrival of the shipment at its destination or of any such shipment that is

lost or unaccounted for after the estimated time of arrival at its destination.

9.6 Lost Material Notification [73.67(e)(3)(vi)]

Intent

The intent of this requirement is to ensure that, in the event a shipment

becomes overdue at its destination or at a scheduled stop on its itinerary and

no reasonable explanation has been received by the licensee from the carrier

regarding its status, a trace investigation will be initiated.

Content

Describe what procedures will be used to trace any shipment that is lost

or has not arrived at a particular point on the shipment's itinerary or at its

final destination by the estimated arrival time.

-5.59-27
9.7 NRC Notification [73.67(e)(3)(vii)]

Intent

The intent of this requirement is to ensure that after a reasonable amount

of time has elapsed with no explanation of the cause for the shipment's being

overdue, the NRC will be immediately notified that the shipment is missing or

unaccounted for and advised of the actions being taken to locate the missing

shipment. The amount of time considered reasonable before these tracing And

notification procedures are initiated should not exceed 1 hour in any case.

Content

Ensure that all material determined to be lost or unaccounted for will be

reported immediately to the appropriate NRC Regional Office in accordance with

§ 73.71. Ensure that notification includes specifying what actions are being

-taken to trace the shipment and that the shipper or receiver, as appropriate,

will also be notified.

5.59-28

.- ;
10.' TRANSFER AND CONTROL REQUIREMENTS

This'chapter provides'guidance on meeting the requirements of

paragraph 73.67(e)(4), which reads as follows:'

(e)(4) Each licensee who'arranges the physical protection of strategic

special nuclear material in quantities of Moderate strategic significance while in transit or who takes delivery of this material free:

on board (f.o.b.) the point at which it is delivered to a carrier

for transport, shall, in addition to the requirements of (e)(1);

(e)(2) and (e)(3) above: -

(i) Make all-shipments of the material either (A) in~dedicated.

.transports with no intermediate stops to load or unload other,,

cargo and with no carrier or vehicle transfers or.temporary

storage in transit, or (B) under arrangements whereby the

custody of the shipment and-all custody transfers are

acknowledged by-signature, and

* (ii) Maintain the material under lock, or under.the.controlof an

individual who has acknowledged acceptance of custody of the

material'by signature.- , - ., . -

10.1 ::Carrier-Transfers (73.67(e)(4)(ij) -. ;--

Intent

.The intent of this requirement is-to ensure.continuity in the assignment

-andiacceptance of personal-responsibility for shipments-ofstrategic special

nuclear material (high-enriched uranium and plutonium). The-requirements of"'.

paragraph-73.67(e)(4) do not apply to LEU. This continuity maybe'provide'd''by'

(l) shipping the material in-a dedicated transport vehicle so that the material

would be under-the continuous control of the same individual(s) for the

duration-of.the shipment or (2) arranging-for each individual who'assumes, fi

custody of-the shipment to-indicate acceptance of personalresponsibility for

the.material by signature. -. , -

Shipment of the material in a dedicated transport means' there' should.be no

intermediate -stops to load or unload other cargo and no,carrier or'.vehicle':.-

transfers or temporary storage en route.. However, material other than'the SNM

being shipped-may be carried in.the transport's cargo compartment'provided it

does not detract from the security~of -the shipment. - . ; , " .

' For shipments by:general freight, this requirement may be satisfied by

arranging for,"signature security- service" with i high.surve'illance:feature,

which is provided by many,.air and motor freight carriers.'

Content

Indicate how this requirement would be met'for''each.of'the'modes'of

shipment to be employed. Indicate how this'requirement'would continue to be

met in emergency situations by reference to the emergency response plans

5.59-29
aincluded in Section 9.4 of the plan. If signature service is to be employed,

describe the type of service that will be provided by the particular carrier

chosen for the shipment. If a dedicated transport is to be used to satisfy

this requirement, affirm that the same individualts) will maintain responsibility for the shipment from the point of origin to its destination.

10.2 Control of Shipments (73.67(e)(4)(ii))

Intent

The intent of this requirement is to ensure that SSNM is not misrouted or

mishandled by unauthorized individuals during the course of a shipment and to

facilitate early detection and recovery of the material in case of loss or

theft. This can be achieved by maintaining the material under lock or under

the control of a responsible individual whenever the material is not secured in

a locked compartment.

Maintaining the shipment in a locked compartment would normally be

possible only for shipments in dedicated transports. In this case, the cargo

compartment should be locked upon embarkation and unlocked only after arrival

at the shipment's destination. If'the locked cargo compartment is left

unattended for an extended period that might allow the unauthorized removal of

the material, the individual(s) responsible for the shipment should establish

upon returning to the transport that the compartment has remained continuously

under lock. This can'be done by examining the cargo compartment lock for signs

of tampering, examining seals on the cargo compartment, or opening the

compartment and ensuring that the material is intact.

' For general freight shipments, during which the cargo compartment may be

opened frequently to load or unload other cargo, locking of. the compartment may

not be practical on a continuous basis. When the shipment is not maintained

under lock, the material'is required to be in a sealed cargo compartment or

maintained under the control:of an individual who has indicated by signature

his/her acceptance of'responsibility for the shipment.- This individual should

be the same person who has signed for custody of the shipment in compliance

with paragraph 73.67(e)(4)(i). For air shipments in which the cargo compartments are not locked, an individual responsible for the shipment is not required

to accompany the shipment during flight provided a responsible individual maintains control of the shipment during the entire time the shipment remains at

each air terminal During unscheduled stops at air terminals'not.on the shipment's itinerary, the shipment may be considered to be under the control of

the flight crew provided the shipment is not offloaded and a member of the

flight crew remains with the aircraft at all times.

Maintaining M the shipment'under control means (1) ensuring the shipment is

properly routed at transfer points, (2) keeping the material under lock or

under surveillance, which would include periodically (normally at least every

2 hours) checking the shipment to ensure it has not been misplaced or tampered

with, and (3) ensuring that the licensee or its designee is notified

immediately if the shipment is determined to be missing. The licensee should

ensure that the signature security service or other service provided by the

5.59-30
carrier under normal or special arrangements includes all three of the these

elements of control.

Content'

Indicate the periods during which the shipment will be maintained under

lock 'and those-during which it will be maintained under the control of a"

'responsible individual. 'Describe the instruments to beused for'documenting acknowledgment of custody of the material by individuals assigned'responsibility for controlling the shipment. Also, describe the arrangements made

with carriers to ensure that the individuals they assign to maintain control of

the material do so in accordance with the three elements of control described

-above- Refer''to the emergency response procedures developed in accordance with

paragraph 73.67(e)(3)(iv) as necessary.

5.59-31
11. EXPORT REQUIREMENTS

This chapter provides guidance on meeting the requirements of paragraph

73.67(e)(5), which reads as follows:

(e)(5) Each licensee who exports special nuclear material of moderate

strategic significance shall comply with the requirements specified

in § 73.67(c), (e)(1) and (e)(3) and (e)(4).

Use Chapters 7, "Material Transportation Requirements," 9, "In-Transit

Physical Protection Requirements," and 10, "Transfer and Control Requirements,"

of Part I of this guide to describe the security procedures that will be used

to protect the material up to the point where the receiver accepts physical

protection responsibility for the shipment.

5.59-32

I I
12. IMPORT REQUIREMENTS

This chapter provides guidance on meeting the requirements of paragraph 73.67(e)(6), which reads as follows:

(e)(6) Each licensee who imports special nuclear material of moderate

strategic significance shall:

(i). Comply with the requirements specified in.5 73.67(c), (e)(2),

;.(e)(3)and (e)(4), and.:

(ii) Notify the exporter who delivered the material to a carrier for

' transport of the arrival of such material. -

12.1 Security Requirements [73.67(e)(6)(i)]

Use Chapters 8, "Receiver Requirements--Transportation,," and 9.

"'In-Transit Physical Protection Requirements,", of Part I of this guide to

describe the security procedures that will be used to protect the material from

the first point at which the shipment is picked up inside the United States.

12.2 Notification'[73.67(e)(6)(iif -

Intent

The intent of this requirement is -to ensure thatthe exporter. is notified

that the material has arrived safely.. .

Content

Describe the procedures to be used for notifying the exporter of the

material that'the shipment has been received..

. .

5.59-33
13. ORDERS TO DELAY SSNM SHIPMENTS

This chapter provides guidance on meeting the requirements of paragraph

73.67(e)(7) and § 73.72 (as it may apply to shipments of special nuclear

material of moderate strategic significance). These requirements read as

follows:

(e)(7) If, after receiving advance notice pursuant to § 73.72 from a

licensee planning to import, export, transport, deliver to a carrier

for transport in a single shipment, or take delivery at the point

where it is delivered to a carrier, special nuclear material of

moderate strategic significance containing in any part strategic

special nuclear material, it appears to the Commission that two or

more shipments of special nuclear material of moderate strategic

significance, constituting in the aggregate an amount equal to or

greater than a formula quantity of strategic special nuclear

material, may be en route at the same time, the Commission may order

one or more of the shippers to delay shipment according to the

following provisions:

(i) The shipper shall provide to the Coauission, upon request, such

additional information regarding a planned shipment as the

Commission considers pertinent to the decision on whether to

delay such shipment.

(ii) The receiver of each shipment, or the shipper if the receiver is

not a licensee, shall notify the Director of the appropriate

Nuclear Regulatory Commission Regional Office listed in

Appendix A by telephone, no later than 24 hours after arrival of

such shipment at its final destination, or after such shipment

has left the United States as an export, to confirm the

integrity of the shipment at the time of receipt or exit from

the United States.

(iii) The Commission shall notify the affected shippers no later than

two days before the scheduled shipment date that a given

shipment is to be delayed.

(iv) Shipments of special nuclear material of moderate strategic

significance which are protected in accordance with the

provisions of §§ 73.20, 73.25, and 73.26 shall not be subject to

orders to delay shipment nor considered to constitute a portion

of an aggregate formula quantity of strategic special nuclear

material for the purposes of determining whether any shipments

must be delayed.

§ 73.72 Requirement for advance notice of shipment of special nuclear

material.

(a) Each licensee who plans to import, export, transport, deliver to a

carrier for transport in a single shipment, or take delivery, at the point

where it is delivered to a carrier, the following materials, shall notify the

5.59-34

K>1_
Director of the appropriate Nuclear Regulatory Commission Regional Office

listed in Appendix A by U.S. Mail, postmarked at least 7 days in advance of the

shipping date: (1) formula quantities of strategic special nuclear material,

or (2) special nuclear material of moderate strategic significance containing

in any part strategic special nuclear material.

(b) The following information.shall be furnished in the advance'notice:

shipper, receiver, carrier(s), estimated dates and time of departure and

arrival, transfer point(s), and mode(s) of shipment.

(c)-' The -Director of the appropriate Nuclear Regulatory CommissionRegional Office shall also be notified by telephone 7 days in advance of the

shipping date'that'an advance shipping'notice has been sent by.mail, and'of any

changes to the shipment itinerary prior to the shipment date. Road shipments

or transfers with one-way transit lines of l hour or less in duration between

installations of a licensee are exempt from the requirements of this section.

-t is noted that shipments of SNM of 'moderate strategic significance.

consisting entirely of uranium enriched in the. isotope. U-235 to 10 percent -or

more but'less than 20 percent are not subject to these requirements.

Of the requirements detailed above, only paragraph 73.67(e)(7)(ii)- and

§ 73.72 should be addressed in the licensee's physical security plan.. :This is

discussed in' Section 13.1.

1. Criteria for DelayinQ Shipments . ;

Under the provisions of paragraph 73.67(e)(7) of the regulations, the NRC

has authority to order delays in certain shipments of SNM of moderate strategic

significance. It is the intention of the NRC to exercise- this authority only

when necessary to protect.the.public health and safety from the possible..

consequences of the theft of a formula quantity of.SSNM. Since the number of

affected shipments expected annually. is quite low, shipment delays are expected

to-be ordered very rarely.. In most cases, licensees will be able to clear.. -

shipment dates with the NRC staff far in advance so that potential conflicts.in

the scheduling of 'shipments can.be avoided.. However, shipment delays may be:

required 'on an emergency basis if a previous shipment is determined to be

missing 'and cannot be'accounted for, before the next scheduled -shipment is to7-

depart. When a scheduling conflict-that cannot'-be resolved voluntarily arises,

the following criteria will be applied to determine which. shipments will be

delayed: .

a. Total time needed to complete the shipment,

b. The' impact'of a- schedule delay on the use of the SNM at the delivery -site, ' ' - . .

c. Planned routingof the material, . -

d. - The relationship of'the material to the national defense or to other

essential programs in the nation's interest, .

5.'59-35
e. The possible impacts on the shipper/carrier relationship with regard

to availability of shipment vehicles and other special arrangements

that would be difficult to reschedule, and

f. How far in advance of the proposed shipment date the NRC was notified K>

of the intent to make the shipment.

2. NRC Requests for Additional Shipment Information

In order to apply the decision criteria listed above, the NRC has the

authority under paragraph 73.67(e)(7)(i) to request additional information from

licensees regarding a planned shipment of SSNM that may be subject to a

possible delay order. Such information requests may be made at any time prior

to a scheduled shipment either by telephone or in writing.

3. NRC Notification of Shipment Delay Orders to Licensees

Under paragraph 73.67(e)(7)Ciii), the NRC will normally notify licensees

that a shipment is to be delayed no later than 2 days before a shipment is

scheduled to depart. In practice, it is expected that potential conflicts will

be resolved much before the'date of a'planned shipment after appropriate NRC

consultations with the prospective shippers. However, in an emergency the NRC

may order delays in shipments or other responsible action at any time it may

deem necessary or desirable to promote the common defense and security, protect

health, or minimize danger to life or property.

13.1 Notification of Receipt of SSNM (73.67(e)(7)(ii))

Intent

The intent of this notification requirement is to ensure that the NRC is

promptly informed of the receipt of a shipment of'SSNM subject to paragraph 73.67(e)(7) or, in the case of export shipments,, informed that the shipment has left the United States safely. Based on this information, the NRC can

permit subsequent shipments subject to paragraph 73.67(e)(7)Cii),to commence

with the knowledge that a formula quantity of SSNM will not be at risk in a

small number of simultaneous shipments. This will also ensure that shipment

delay orders will not have to be imposed unnecessarily for failure to confirm

that a previous shipment is no longer at risk. Compliance with this

requirement is not a substitute for compliance with other notification

requirements under paragraph 73.67(e)(2)(ii).

Content

Affirm that, in the case of a shipment of SNM of moderate strategic

significance containing in any part SSNM, arrangements are made to obtain

confirmation of the integrity of the shipment at the point where the material

leaves the United States as an export or at its final destination. Describe

the procedures for accomplishing this and notifying the NRC of the shipment's

status at the time of receipt or.exit from the United States. Include the

names or titles of individuals who will be responsible for ensuring that the

required confirmation is made and the proper notification is provided to the

NRC.

5.59-36

<1112
PART II

SPECIAL NUCLEAR MATERIAL

OF LOW STRATEGIC SIGNIFICANCE

5.59-37
1. USE AND STORAGE AREA AT A FIXED SITE

This chapter providesRguidance on meeting-the requirement of paragraph

73.67(f)(1), which is as follows:

(f)(1)Y. Store or use such material only within a controlled access area.

AI controlled access area (CA) is defined in paragraph 73.2(z) as "any

temporarily or permanently established area which is clearly demarcated, access'

to which.s controlled and which affords isolation of the material or persons

within it.". "Access control"'means measures used to allow only specified

personnel, materials, and vehicles ingress-intoxand:egress from a given area,

while "isolation" refers to measures taken.to deter persons, materials, or

vehicles from entering or leaving a given area through other than established'

access-control points. Thus, a CAA includes-provisions for both isolation and

accesscontrol. In some cases, isolation or. access control systems may also'

serve the purpose of aiding in the detection of unauthorized penetration or

activities* within the CAM. Therefore, these detection-related considerations

-are also-alluded to in this chapter.

In the discussion that follows, CAAs intended for the use and possible

storage of SNM of low strategic significance are discussed separately from

those intended solely for the storage of such SNM. Although there is no

difference in the requirements for these two different applications of: the"CAA,

it is recognized that the means used to isolate the material and to control

access may differ markedly for use areas compared with storage areas and

therefore warrant separate discussion.

1.1 Areas for Use and Temporary.Storage [73.67(f)(1)]

,.,This section discusses.CAAs intended primarily.for the use and temporarystorage of SNM of low strategic significance. These may be temporarily established to meet-transitory or intermittent SNM use requirements or-they may bepermanently established. Permanently.established CAMs for use of. SNM may.also

be used'-for storage.for short intervals between periods of usage.. "Use" -means

that the material is undergoing processing (e.g., fuel fabrication, irradiation

in a reactor) or utilization of its properties in conjunction with experimental

equipment (e.g.,'equipment used for research or educational laboratory

experiments). Different isolation/access-control measures may be used for

periods during which'the'area is occupied versus unoccupied..

1S Temporarily Established CAAs . -

Temporarily'established CAAs for the use of. SNM need not have permanent

barriers at their boundaries. Isolation of the material and persons using the

material may be provided by office partitions, cordons, or other devices used

to warn passersby of the restricted nature of the, area. Access control can be

effected through surveillance or 'supervision of the area by those who are using

the SNM and who are responsible for the material. However, the provision of

Unauthorized activities are those activities deemed by the licensee to.be -

indicative of or contributory to the possible theft of"SNM. '

5.59-39
access control through personal supervision is suitable only for those

situations in which the size of the CAM and the number of persons to be

admitted are sufficiently small and in which the CMAis'suitably configured to

make such procedures practical.

When material located in a temporarily established CAM is to be left

unattended (i.e., because it is impractical to replace the material in a CAM

designed for long-term storage of the material), access control and'improved

isolation for the temporary CAM may be provided to substitute for the discontinued personal supervision over the material. (In addition, provisions must

also be made for satisfying the monitoring requirement of paragraph 73.67(f)(2).)

Improved isolation can sometimes be provided by increasing the penetration

resistance of existing barriers (e.g., locking doors and windows, placing the

material in a locked drawer or supply room). Access control may be provided by

any of the measures suggested below for permanently established CAAs left

unattended; some of these suggestions may be more suitable for temporary use

than others. If no suitable barrier is available to provide isolation, material

left unattended may be protected by a motion alarm covering'the area immediately

surrounding the material. This improved detection capability would be considered an acceptable substitute for the rudimentary isolation provided by

temporary barricades, cordons, signs, and other less substantial means of

isolation that may be used on a temporary basis.

2. Permanently Established CAAs

Permanently established CAAs for the use and temporary storage of SNM of

low strategic significance would most likely provide isolation for the SNM

through the use of permanent barriers. These could consist of fences; gates or

freestanding walls for exterior areas; or exterior or interior building walls,

locked doors, windows, bars, grillwork; or other barriers for interior areas.

Such barriers are not required to meet the more stringent-criteria for physical

barriers used for the protection of formula quantities of strategic special

nuclear material (SSNM).* However, good security management practice would

dictate that the barrier be substantial enough to deter casual passersby from

unauthorized penetration. If the barriers are also designed to aid in detection, the criteria for penetration resistance or tamper indication in accordance with the specific monitoring procedures to be used also apply.

Access control for permanently established CAMs during periods they are

occupied can be provided by personal supervision. However, when these areas

are unoccupied or when the size, configuration, or numbers of persons intended

to be admitted to the area make personal supervision impractical, other means

of access control may be called for. Some of the additional access control

measures that may be used for permanently established CMAs in these situations

are:

Stationing a watchman at CAM access control points,

Limiting distribution of keys, keycards, or combinations to locks on

doors and gates,

AuStrategic special nuclear material" means uranium-235 (contained in uranium

enriched to 20 percent or more in the U-235 isotope), uranium-233, or plutonium.

5.59-40
Using a coded badging system to identify authorized personnel. at CAA

access control points,

Controlling locked CAA doors and gates by use of remote surveillance

(e.g., intercom or CCTV) by personnel stationed at a centrally.

located access control facility, or ,

'Controlling access of personnel, material, 'and vehicles into CAAs by

other means. -

-'It is expected that large facilities (e.g. fuel processing facilities)

will have more complexisolation and access control systems than small facilities (e.g., research facilities) in order to achieve. comparable levels of

protection.

Content'

Describe the CAAs where the material will be used. Indicate under'what

conditions CAAs will be established on a temporary basis. State whether each

CAA will also be usedfor'storage." For'each CAM identified,''provide the

following information: '

1. A description of the area and its features relative to other facility

features, 'showing the normal routes of ingress'to each CM (including a scale

diagram). '

2. Descriptions of physical barriers, cordons, walls, partitions, or

other means of providing isolation of the CAM, and channelling entry through

established access control points into the CAM.

3.' 'The means and criteria used for controlling access into the CAM at

established access control points. ' '

4. If more than one CAA is designated, the types of material normally

used in each CAM; for temporarily established CMs, the types of activities

normally performed within each area.

'1.2 -Areas for Permanent Storage [73.67(f)(1))

Intent

CAMs normally used only for storage of SNM are described in this section.

Such CAMs may be very limited in size since they need only contain the material

itself, and access to them-is expected to be infrequent. ' Where'small amounts

of material are involved, the C M may consist of-relatively small containers

such as security cabinets, safes, and locked closets or supply rooms.

Isolation for CAMs intended 'onlyfor storage usually'will be provided by

the penetration-resistant features of the perimeter barriers or'container walls'

of the CAM. Such barriers and container'walls-are not required to meet the

more stringent criteria for physical barriers used for the protection'of

formula quantities of SSNM. However, good security management practice would

5.59-41
dictate that the barrier be substantial enough to deter casual passersby from

unauthorized penetration. If the barriers are also designed to aid in

detection, the criteria for penetration resistance or tamper indication in

accordance with the specific, monitoring procedures to be used also apply.

In determining the level of isolation and access control for such CAAs,

the minimum time required for extricating SNM from its storage location in

relationship to the licensee's time. of detection and assessment may need to be

taken into account (e.g., fuel assemblies stored in shipping casks or

non-self-protecting fuel residing in nonpower reactor cores). The bulkiness,

massiveness, and difficulty of movement or disassembly of fresh fuel elements

or other forms of low-enriched uranium (LEU) in large quantities such as

shipping casks or assemblies may pose sufficient difficulty to persons attempting unauthorized removal within a reasonable time period (relative to CAM

monitoring procedures) that they may be considered to provide isolation.

Control of access to special equipment and vehicles capable of handling and

transporting such items may also be included as integral elements of the access

control program for the CAM.

Access control procedures for CA s designated exclusively for storage are

expected to be similar to those described in the previous section for periods

when use-type CAAs are unoccupied.

It is expected that large facilities (e.g., fuel processing facilities)

will have more complex isolation and access control systems than small facilities (e.g., research reactors) in order to achieve comparable levels of

protection.

An illumination level is'recommended for the CAM that is sufficiently

uniform and bright to detect penetration of or tampering with the CAM containment (in the case of a security cabinet) or unauthorized penetration of or

activities within the CAM (in the case of a vault-type room or its equivalent).

Content

(This section needs to be completed only for CAAs designated exclusively

for storage of SNM. CAMs designated for both use and storage are to be

described as recommended in Section 1.1 of Part II of this Standard Format.)

Describe the CAMs where the material will be stored, if'different from

those previously described for use. For each such CAM, provide the following

information:

1. A description of the area and its features relative to other facility

features showing the access control points for the CAM (including a scale

diagram).

2. The physical barriers, container walls, or other features that

demarcate the perimeter of the CAM and channel entry through only established

access control points. If access control is provided by virtue of the

bulkiness, massiveness, and difficulty of disassembly of a container or object

5.59-42
serving as a CAM, describe the facts relied upon to support such a

determination.

3. The means and criteria used for controlling access to the CAM at

established access control points.

4. If more than one CAM is-designatedrforithe storage of the material,

the types of material normally to be stored in each area.

.I

I

'I:

5.59-43
2.. DETECTION DEVICES AND PROCEDURES AT A FIXED SITE

This chapter provides guidance for meeting the requirement of paragraph 73.67(f)(2), which is as follows:

(f)(2) Monitor with an intrusion alarm or other device or procedures the

controlled access areas to detect unauthorized penetrations or

activities.

The purpose of this monitoring activity as stated in paragraph 73.67(a)(2)

is to provide "early detection and assessment of unauthorized access or activities by an external adversary within the controlled access area..." and "early

detection of removal of special nuclear material by an external adversary from

a controlled access area." This should be done to achieve the objective of

minimizing the possibilities for unauthorized removal of special nuclear

material "consistent with the consequences of such actions," as stated in

paragraph 73.67(a) of the effective rule. Thus, the earliness of detection

afforded by the licensee's physical protection system may vary depending on the

nature and quantity of the material susceptible to unauthorized removal.

Further discussion of the earliness of detection required for possible thefts

of different types and quantities of special nuclear material of low strategic

significance is provided in Section 2.1 of Part II of this guide.

2.1 Earliness of Detection (73.67(a)]

The criteria used for determining the earliness of detection for unauthorized access, activities, or removal of SNM from CAAs will depend on an assessment of the magnitude of the consequences associated with possible misuse of

the type and quantity of material that could be removed in a given theft

attempt. For SNM of low strategic significance, three distinct cases are

considered:

1. Gross theft of low-enriched uranium (LEU)*

2. Minor theft of LEU.

3. Theft of strategic special nuclear material (SSNM) (i.e., SNM of low

strategic significance other than LEU).

Gross theft of LEU refers to the theft of LEU in a sufficiently large

quantity that it could yield upon further enrichment or other processing enough

material of the type and quantity needed to construct a clandestine fission

explosive (CFE) device. This quantity may vary depending on the enrichment and

physical and chemical form of the feed material but may roughly be estimated as

an amount containing about 75 kilograms of the isotope U-235. Because the most

serious consequences to the public health and safety could occur from a theft

of this magnitude, a system that would detect a theft sufficiently early to

allow for prompt recovery is called for. The NRC believes that, because of the

*

Low-enriched uranium is uranium enriched above natural but less than 20 percent

in the U-235 isotope.

5.59-44

K>1
large quantity of material involved and with:proper detection systems and

procedures' the licensee'would be capable of detecting a gross theft of LEU-:

between the time of the actual theft and the time it is transported off site.

'Minor. theft ofLEU refers to thefts involving much smaller quantities of

LEU such as could'be removed by one or.,two persons in a private vehicle or on

one's person. ..These minor thefts-are.significant only to the extent that they

-could'be.repeated periodically to eventually accumulate an-aggregate~quantity

similar to'that which might be obtained, in a gross theft.- The required degree

of.-earliness -of detection of such thefts is related to.the time that would'be

needed-by the adversary to'obtain a gross amount through repeated thefts. If.

the -amount taken..in each of a series of daily thefts is very small, the -

ultimate6time of detection could be as long as 6 months--.the interval between

material .inventories 'as required by Part 70. For larger quantities that could

be removed in a single theft, but still much less than would constitute a gross

theft, it'woul d be reasonable 'to expect that one in a series of such thefts

would be detected before a gross amount could be removed.,

Licensees who possess less than gross quantities of.LEU need not provide

for early detection of a s'ingle theft-of a gross quantity,2but the capability'

for detecting multiple thefts is as important as it is for those. licensees - possessing gross' quantities. ' :

Thefts from a single facility of SSNM (SNM of low strategic significance

other than LEU) are limited by definition -to quantities that could not be used

-to construct a nuclear. explosive device; thefts of similar quantities of.SSNM

'from several different facilities'could, however, lead to the -accumulation of

an 'aggregate quantity that would permit such illicit use...(The~consequences of

a single theft of such material would have minimal impact on the public health

and safety.) Therefore, detection of suchthefts is-required.sufficiently

early to ensure'early notification to the NRC. The NRC in turn can then notify

,other licensees of the need to implement appropriate response measures to

prevent the accumulation by'a single adversary, of. a formula -quantity.of material through multiple thefts from'different facilities. The NRC believes

that there are detection systems and procedures that-would allow thelicensee

to detect a theft of SSNM within approximately 2 hours. , -

2.2 Monitoring Controlled Access Areas [73.67(f)(2)]

Intent . . .

Monitoring CAAs where the material is used will require a- capability for

early detection of (1) unauthorized penetration'of the CAM perimeter,'(2)' --unauthorized activities within-the CAA, (3) unauthorized removal of SNM from

the CMA, or'(4).some combination-of these, 'by'an external adversary. ' For -

purposes of this guidance, an-external adversary is any person-not employed by

the licensee or one of its contractors to report for work on a regular basis at

the licensee's site and who.is not.directly involved in the processing or other

authorized use of the material in connection with'his/her assigned duties. For

purposes of this definition' graduate and'other students assigned by the

licensee at an educational or research institution to assume responsibility for

5.59-45
the regular handling of the material would not be considered an external

adversary.

The choice of particular safeguards measures to be included in the CAM

monitoring system will depend on a number of different site-specific factors

such as whether the measure is to function while the-area is occupied or

unoccupied, the size of the area, the types of functions'performed in the area,

the level of traffic into and out of the;area, etc. Many of the available

safeguards measures that may be employed in the CAM monitoring system are

described in various regulatory guides and NUREG documents issued by the NRC.

Specifically, the Fixed Site Physical Protection Upgrade Rule Guidance Cortpendium* provides guidance on selection, installation, and implementation of

monitoring devices and procedures relative to physical protection of formula

quantities of SSNM. This guidance is also applicable to monitoring equipment

and procedures for CAAs at fixed sites having less than formula quantities of

SNM with regard to individual physical protection measures. It is recognized,

however, that from a systems point of view, this body of guidance'intended to

support the Upgrade Rule may emphasize a much greater level of redundancy and

diversity than would be required for the protection of SNM of'moderate and low

strategic significance.** The following paragraphs give illustrations of

various monitoring systems that can be used to provide the earliness of detection discussed in Section 2.1 of Part II of this guide.

1. Early Detection of Gross Theft of LEU

Early detection of gross theft of LEU means detection during the attempted

theft. This may be accomplished by using intrusion alarms or other monitoring

devices in the following applications:

Fence or buried line perimeter detection systems'at the C M boundaries.

Door and window alarms (e.g., balanced magnetic switches) for CAMs within buildings and alarms on emergency exit doors equipped with crash bars.

Volumetric interior motion detection systems for unoccupied'CAMs or

portions of CAMs within buildings.

In many cases, security procedures may be employed as substitutes for some

or all of the devices that could be employed to fulfill the ponitoring

requirement for the CAM. Security procedures may be especially attractive in

protecting against the theft of gross quantities of LEU located outside

The compendium was issued in conjunction with the Physical Protection Upgrade

Rule. It is intended to assist the licensee in theldevelopment and implementation of safeguards physical protection and transportation protection plans.

Copies are available for inspection at the Commission's Public Document Room,

1717 H Street NW., Washington, D.C.

Tamper-safing of devices intended for use at facilities having less than

formula quantities of SSNM need protect only against the external'adversary;

to the extent such a distinction can be made for a given'device, this

represents a relaxation in tamper-safing requirements relative to protection

of formula quantities.

5.59-46
buildings where intrusion alarms may more easily be circumvented by an external

adversary. Typical.security procedures that may be employed for detection of

attempts at unauthorized removal of gross quantities of LEU include the

following:

Periodic patrols of CM perimeters and areas within the CAM by

watchmen.

' Surveillance (escorting) of nonlicensee vehicles capable.of transporting gross quantities of LEU away from the site.

Continuous (without interruption) or continual (intermittent)

surveillance of the CAA, or. a given portion of the CAAMthat contains

a gross quantity of LEU, by designated supervisory or other licensee

employees. (These surveillance responsibilities could be undertaken

in addition to the employee's normal functions provided they can both

be.discharged adequately.).

* Periodic inspections to confirm continued integrity of barriers,

gates, loading. bay..doors, etc., through which unauthorized removal -of

gross quantities of LEU could be effected. -

' Inspection of large vehicles leaving CAAs to ensure they are not used

for unauthorized removal of a gross quantity of LEU.

Where procedures involve-periodic patrols-or inspections,'the amount of

time that would be required for an adversary to complete the removal'of a gross

quantity of LEU may be used to govern the interval between inspections. -For

example, if a given number.of a certain type of container of LEU would have to

be-individually loaded onto a-tiuck in order to effect removal of a gross

quantity of LEU and the' time required.for'.such loading is determined toube in

excess of 2 hours, then a 2-hour guard patrol of the given area would be

considered to meet the earliness-of-detection guideline for the subject

material. -

A 'combination of these and perhaps other measures could satisfy the early

tdetection criterion'for gross theft of 'LEU if it could be 'shown that an.

external adversary could not obtain unauthorized access to the SNM without

being detected by at least one intrusion alarm component.

2.. Early Detection'of Minor Theft of LEU;.

a The criteria for early detection of minor theft of LEU can be satisfied by

a monitoring program that provides 'a'sufficiently high probability of detection

of one in aseries of minor thefts so 'itis implausible that an adversary would

be able to obtain in'the aggregate-a gross quantity of LEU.' Such a capability

could be provided'in whole, or'in part, by'thesame monitoring program designed

to detect gross-theft. Additional measures that might be included in the

monitoring program to protect against minor theft include the following:

Exit searches of nonlicensee vehicles capable of transporting

hand-carried'quantities of LEU (on a'random or general basis).

5.59-47
Administrative control of materials removed from controlled areas

where easily concealed or hand-carried items containing LEU are

available.

Item count inventories more frequent than those currently required

under Part 70.

Procedures directing regular employees to be aware of unauthorized

persons in their work areas and to report the presence of such

persons to supervisory or security personnel. (A coded badging

program would support this measure.)

Escorting of visitors and other unauthorized persons entering CAAs.

Watchman patrols to detect unauthorized persons within the CAM.

* SNM doorway monitors at all personnel access control points.

These are not required measures but illustrations of how the monitoring

program for detecting gross theft of LEU may be extended to'include the capability for early detection of minor theft of LEU. One or more of these measures may be employed, but probably not all of them would be needed for any one

installation.

3. Early Detection of Theft of SSNM (SNM of low strategic significance

other than LEU)

For facilities where it is necessary to protect against thefts of SNM of

low strategic significance other than LEU (SSNM), earliness, of detection is

discussed in Section 2.1 of this chapter. The physical protection system in

this case may differ considerably from that designed to protect LEU. The sites

are likely to be small research-oriented facilities such as nonpower reactors

and educational institutions. Also, there is likely to be much less material

available, and authorized access to it may be less frequent. Detection devices

such as interior motion detector systems, balanced magnetic switches, etc.,

could be used to monitor relatively small CAMs where the number of established

access control points is very limited.

Procedures may also be used to protect against thefts Of SSNM in this

subcategory. Since watchman patrols inherently allow unguarded intervals

during which an external adversary could conceivably obtain access to the CAM,

remove a small amount of material, and leave before being discovered, the

inspection procedure may (a) be sufficiently frequent to ensure an external

adversary could not complete a successful theft of the material during the

unguarded intervals,' (b) provide for an irrevocable indication of the CAA

barrier having been tampered with (as with a seal or by observation of damage

done to the barrier)', or (c) provide for an item count inventory or other

accounting for the material as part of the inspection. Some examples are given

below:

After having left'material unattended in an experimental setup in a

classroom used as a temporary CAM, the instructor returns after

2 hours and verifies that the material is still in place.

5.59-48

<2
A small quantity of material is stored.in a security cabinet for

several months at a time without authorized.access. Every 2 hours a

watchman comes around to inspect the cabinet to ensure that the lock

is secure and the cabinet has not been tampered with. If a suitable

combination or other pick-resistant lock is not used, the watchman

also inspects a tamper-indicating seal to verify that the container - has not been 'tampered with.

During periods when a nonpower reactor is left unattended, a watchman

comes around every 2 hours and checks that none of the fuel elements

have been removed by unauthorized persons. If a person'working during

off-hours in the 'vicinity of-the reactor is encountered, appropriate

identification is requested.

While a nonpower reactor is being used, the supervisor ensures that

he/she or his/her designee is not away from the'reactor-for a period

longer than 30'minutes, a period he/she estimates would be minimally

required for a person to secure fuel elements in the core that were

not self-protecting by virtue of their having been sufficiently.

irradiated.

.'During periods when excess fuel is left unattended in a storage

closet at a nonpower reactor~facility, an alarm system is activated

that would annunciate if anybody tried to penetrate the closet

without proper authorization.

During periods when the material is being used under the direct supervision of authorized personnel, the monitoring requirement is satisfied by virtue

of the continuous surveillance by the authorized personnel.

Content

For each of the CAAs described previously (see Chapter 1 of Part II of

this guide), provide the following information regarding the measures taken to

fulfill the monitoring requirement of paragraph 73.67(f)(2):

1. Describe any devices employed by specifying the type of device, its

installed location, the type and location of annunciation, its intended area of

coverage, and its tamper-resistant features. Refer, if desired, to existing

filed security plans, regulatory guides, or NUREG documents.

2. Describe any procedures employed to monitor the CAM or portions

thereof, including the categories of persons who will execute the procedures,

the frequency of inspections or rounds of patrol, the basis for the determination of such frequencies, and the occasions upon which the procedures are

intended to be implemented. Also indicate the features of the CAM (e.g.,

barriers, locks, seals) that will play a role in the way the procedures are

used.

3. Explain how the combination of procedures and devices used to monitor

each CAA meets the criteria for early detection of theft by an external

adversary relative to the type of material found in each CAA.

5.59-49
3. SECURITY RESPONSE AT A FIXED SITE

This chapter provides guidance on meeting the requirement of paragraph

73.67(f)(3), which is as follows:

(f)(3) Assure that a watchman or offsite response force will respond to all

unauthorized penetrations or activities.

Intent

The intent of this requirement is to ensure that, in the event of a

security incident, someone will be available to assess alarms or any unauthorized penetrations or activities and, if warranted, notify the NRC, the local

law enforcement authorities, and the responsible person in licensee management.

Early detection and notification of any missing material will help facilitate

its prompt recovery. For the purpose of this regulation, an offsite response

force can be a local law enforcement agency or a contract guard service.

Content

Describe the security organization that will be responsible for assessing

and responding to any unauthorized penetrations or activities. Ensure that at

least one guard, watchman, or member of an offsite response force will respond

to all unauthorized penetrations or security incidents at the CAAs.

5.59-50
4. RESPONSE PROCEDURES AT A FIXED SITE

This chapter provides guidance on meeting'the requirements of paragraph 73.67(f)(4), which states:

(f)(4) Establish-and maintain response procedures for dealing with threats

of thefts or thefts of such material.

Intent

) ':

-The intent of this regulation is to help the licensee to identify those

security-incidents that could result in the,,loss',of, SNM of low strategic

significance and to develop response procedures to prevent 'or reduce the

likelihood of such a loss. Some types of incidents that should be considered

and for which-response procedures should be developed'are:

1. Situations that could possibly lead to theft of SNM (e.g., civil

disturbance).

2.- Discovery that the security system has been breached.

3. Discovery~that some' SNM is missing.

Content

Identify'those events for which response procedures will be developed.

Also describe the type of response'to' be accomplished for each event identified

and the duties and responsibilities of the security organization and management

involved in the response. Ensure that the NRC will be notified immediately in

the event of theft or'attempted theft of the material. Describe what local law

enforcement assistance is available, their response capabilities, and any . ,

agreements made with them to respond in the case of theft of the material. -

..

. , . . I ..

5.59-51
5. MATERIAL TRANSPORTATION REQUIREMENTS

This chapter provides guidance on meeting the requirements of paragraph

73.67(g)(1), which are as follows:

(g)(1) Each licensee who transports or who delivers to a carrier for

transport special nuclear material of low strategic significance

shall:

(i) Provide advance notification to the receiver of any planned

shipments specifying the mode of transport, estimated time of

arrival, location of the nuclear material transfer point, name

of carrier and transport identification,

(ii) Receive confirmation from the receiver prior to commencement of

the planned shipment that the receiver will be ready to accept

the shipment at the planned time and location and acknowledges

the specified mode of transport,

(iii) Transport the material in a tamper-indicating sealed container,

Civ) Check the integrity of the containers and seals prior to shipment, and

(v) Arrange for the in-transit physical protection of the material

in accordance with the requirements of § 73.67(g)(3) of this

part, unless the receiver is a licensee and has agreed in

writing to arrange for the in-transit physical protection.

5.1 Advance Notification [73.67(g)(1)(i)]

Intent

The intent of this paragraph is to require the shipper to preplan the

transportation of the material and inform the receiver of his plans prior to

shipment. This is the first of several transportation requirements that will

allow the receiver to take delivery of the material as planned or to help

ensure traceability of any missing material.

Content

The licensee should ensure that, prior to each shipment of material, the

receiver will be notified of the impending shipment and provided the following

types of information:

1. Mode of transport (e.g., truck, plane, train, or ship),

2. Estimated time of arrival,

3. Location where material is to be transferred to receiver,

4. Name of carrier, and

S. Transport identification (e.g., truck, train, or flight number; ship

name).

5.59-52
5.2 Receiver Confirmation [73.67(g)(1)(ii))

Intent

The' intent of this requirement is that, prior to shipment, the transporter

will be assured that the receiver is ready to accept the shipment at the

planned'time'and location'and has acknowledged the mode of transport.

Content

Describe what procedures will be used to ensure that shipment of material

does not take place until the'receiver'acknowledges the planned shipment and:'

mode of transport and readiness to accept the shipment at the planned time and

location.

5.3 Container C73.67(g)(1)(iii)]

Intent

The'intent of this requirement is to provide a mechanism or system that

will help'the receiver detect any tampering with the material's container that

may have occurred during shipment. Regulatory.Guide 5.15, "SecurityfSeals for.:

the Protection and Control of Special Nuclear Material," provides guidance in

this area. If the material is shipped in an exclusive-use carrier, it is

acceptable to tamper-seal the carrier itself rather than each individual SNM

container.

Content

Describe the types of seals that will be used to secure the material's

container during transport.

5.4 Inspection [73.67(g)(1)(iv)]

Intent

The intent of this paragraph is to require the shipper to check the

integrity of the material container's seals just prior to shipment so as to be

assured that they have not been compromised. Then, upon receipt of the

shipment, if the receiver discovers the container's integrity has been

compromised and the material is missing, the scope of the recovery operation

can focus on the transportation route.

Content

Describe the procedures to be used to ensure that the integrity of the

containers or seals is checked just prior to shipment.

5.59-53 .
5.5 Responsibility for In-Transit Physical

Protection L73.67(g)(1)(v))

Intent

The intent of this paragraph is to make clear that the licensee shipping

the material is responsible for arranging for the physical protection of the

material in transit if the receiver is not a licensee. If both the shipper and

receiver are licensees, the shipper may allow the receiver to accept this

responsibility wholly or in part provided there is appropriate documentation

specifying their respective responsibilities. Where no such documentation

exists, both the shipper and receiver are held jointly responsible. (See

Section 6.3 of Part II of this guide.)

Content

In its security plan, the shipper should either acknowledge responsibility

for the in-transit physical protection of SNM of low strategic significance or

ensure that a written agreement from the receiver has been received in which

the receiver accepts either full responsibility or shared responsibility for

the in-transit physical protection of this material in accordance with

paragraph 73.67(g)(3) of 10 CFR Part 73.

5.59-54

K>
6. RECEIVER REQUIREtIENTS--TRANSPO.TATION ,

This chapter provides guidance on meeting the requirements of paragraph

*73-67(g)(2), which are as follows:

(9)(2) Each licensee who receives quantities and types of special nuclear

material of low strategic significance shall: -.,

(i) Check the integrity of the containers and. seals upon receipt of

the shipment,

. -(ii) Notify theshipperof receipt of the material as required in

; § 70.54 of Part 70 of this chapter, and

(iii) Arrange for the in-transit physical protection of the material

in accordance with the requirements of paragraph 73.67(g)(3) of

this part, unless the shipper is a licensee and has agreed in

writing to arrange for the in-transit physical protection.

6.1 Inspection [73.67(q)(2)(i)]

Intent

This requirement is intended to determine whether the material's container

has been compromised en route and whether any material has been removed so that

immediate recovery procedures can be initiated if required.

Content

Describe the procedures to be used to ensure that the integrity of the

containers and seals will be checked upon receipt of the material shipment.

6.2 Notification (73.67(g)(2)(ii)]

Intent

This requirement is intended to ensure that knowledge of the current

location of all SHM is available and to formally inform the shipper that the

material has been received. %

Content

Ensure that a completed copy of Form NRC-741, "Nuclear Material Transaction Report," will be sent to the shipper within 10 days-after a material

shipment has been received as required in § 70.54 of 10 CFR Part 70.

6.3 Responsibility for In-Transit Physical

Protection (73.67(q)(2)(iii)]

Intent

The intent of this paragraph is to make clear that the licensee receiving

the material is responsible for arranging for the physical protection of the

5;59-55:
material in transit if the shipper is not a licensee. If both the shipper and

receiver are licensees, the receiver may allow the shipper to accept this responsibility either wholly or in part provided there is appropriate documentation

specifying their respective responsibilities. Where no such documentation exists,

both the shipper and receiver are held jointly responsible. (See Section 5.5

of Part II of this guide.)

Content

In its security plan, the receiver should either acknowledge responsibility for the in-transit physical protection of SNM of low strategic significance

or ensure that a written agreement from the shipper has been received in which

the shipper accepts either full responsibility or shared responsibility for the

in-transit physical protection of this material in accordance with paragraph

73.67(g)(3) of 10 CFR Part 73.

5.59-56
7. IN-TRANSIT PHYSICAL PROTECTION REQUIREMENTS

This chapter provides guidance on meeting.the requirements of paragraph

7 3 .67(g)(3), which are as follows:

(g)(3) 'Each' licensee, either shipperor receiver, who arranges for the

physical protection of special nuclear material of low strategic,

significance while in transit or who takes delivery of such material

free on board (f.o.b.) the point at which it is delivered to a

carrier for'transpbrt shall:

* "'. (ij -Establish and maintain response procedures for dealing with

'' ' 'threats of thefts or thefts of such material,

(ii) Make arrangements to be notified immediately of the arrival of

* the shipment.at its destination, or of any such shipment that is"

lost orunaccounted for after the estimated time of arrival at

'its destination, and . ,

(iii) Conduct immediately a trace investigation of any shipment that

is lost or unaccounted for after the estimated arrival time andreport to the Nuclear Regulatory Commission as specified in'

§ 73.71 and to the shipper or receiver as appropriate..- The

-' licensee who made'the physical protection arrangements shall

* also'immediately notify the'Director of the appropriate Nuclear'

Regulatory Commission Regional Office listed-in Appendix A of

..the action being taken'to trace the shipment..

7.1 Response Procedures'[73.67(g)(3)(i))

Intent

The intent of this regulation is to help the licensee identify those

transportation incidents that could affect the security of the SNM in transit

for which notification might be expected and for which response procedures

should be planned.

Content

Identify those events for which response procedures will %be developed.

Also describe the type of response to be accomplished for each event identified

and the duties and responsibilities of the security organization and management

involved in the response. Ensure that the NRC will be notified immediately in

the event of theft or attempted theft of the material.

7.2 Notification r73.67(g)(3)(ii)]

Intent

The intent of this requirement is to ensure that the licensee responsible

for the physical protection of SNM in transit will have a firm basis for

deciding whether or not to initiate response procedures in the event a shipment

becomes overdue or is lost.

5.59-57.
Content

Describe the-arrangements and procedures that will be.used for notifying

the licensee who arranges for the physical protection of material in transit

(1) of the arrival of the shipment at its destination or (2) of any such

shipment that is lost or unaccounted for after the estimated time of arrival at

its destination.

7.3 Lost Material Notification [73.67(g)(3)(iii)]

The intent of this requirement is to ensure that, in the.event a shipment

becomes overdue and no reasonable explanation has been received from the

carrier regarding its status, a trace investigation will be conducted to locate

the missing SNM. At this time, the NRC should be notified that the material is

missing and informed as to what steps are being taken to recover it. Although

the licensee is responsible for notifying the NRC of any missing material and

for initiating and assisting in the subsequent investigation, the law

enforcement agencies bear the responsibility for physically recovering the

material.

Content

Describe what procedures will be used to trace~any shipment that is lost

or has not arrived by the estimated arrival time. Ensure.that all lost or

missing material will be immediately reported to the appropriate NRC Regional

Office along with what actions are being taken to trace the shipment, that the

NRC will be notified as specified in § 73.71, and that the shipper or receiver,

as appropriate, will also be notified.

5.59-58
Appendix 12.4

Regulatory Guide 2.5

Quality Assurance Program Requirements

for Research Reactors

STANDR~D FoRmAT AmD CoNT~Eb-T
U.S. NUCLEAR REGULATORY COMMISSION Revision O-R

lotMay 197

5) REGULATORY GUIDE * ' OFFICE OF STANDARDS DEVELOPMENT (This guide reissued

October 1977)

REGULATORY GUIDE 2.5

QUALITY ASSURANCE PROGRAM REOUIREMENTS FOR RESEARCH REACTORS

A. INTRODUCTION

Paragraph (aK7) of §50 34. "Content% or Applica- tion%. Technical Information or to Cc-R P.rt 50.

*Licening of Production .ind Utilwition I J.111ticht

requires thit eich applicant for a contruttion permit

to build a production or utilization facilit inmludc in

it% preliminar) %afet) anjalis report a deLription ol

the quilit) j .iurance program to be applied to the

design and con truction of the %tructure%. %)%temr

and component% of the lacilit, Paragraph (b)6Xit)

or 50 34 reqwre. that each ipplitint for a. bcense t)

operate a facilit) include in the final %alct) inal.i.

report a description of the managerial and admmistrativc controls to be used to assure safe operatlion This guide describes a method acceptable to the

NRC staff of complying with the Commision's

regulations with regard to overall quality assurance

program requirements for research reactors

B. DISCUSSION

Work Group ANS-15 8 of Subcommittee ANS-15.

Research Reactors. of the American Nuclear Society Standards Committee has developed a standard that

describes a quality assurance program for use in

research reactors The standard was approved by the

American Nationil Standards Committee N17.

Research Reactors It wSV subsequently approved

and designated ANSI N402-1976 b) the American

National Standards Institute (ANSI) on August 19.

1976

C. REGULATORY POSITION

The gcneral requirements for establishing and exuting .i qu.lts avurance program for the demtgn.

Rtontrut.tion tctinL modification. and mintenance

tit re erah reictorN that arc induded in ANSI N402-

1976 "Qu.lilth A%%urince Program Rcquirement% or

Reoai,.h Rcator% provide an acceptable method

lor Lomphing vith the program requirements or 10

(I R ki(l U-3

D. IMPLEMENTATION

The pulpo..c ol thi% 'ection i% to provide information to applicants regarding the NRC starrb plans for

uwang thi' regulator% guide

Except in tho e cIsCs in which the applicant

propo e' .n acceptable alternative method for compl ing 'Ath 'apecified portions of the Commission's

regulationk. the method described herein wtll be used

in the evaluation of 'aubmittals in connection with applications for operating license%. construction per- mits, or propo ed amendments thereto docketed after

January 16. 1978. unless this guide is revised as a

result of suggestions from the public or additional

'trt reviecA

If an applicant vishes to use this regulatory guide

in developine ubm itt.tls for applications docketed on

or before January 16. 1978. the pertinent portions or

the application will bt evaluated on the basis of this

guide

'Lopoeb mat be ahijrned Irom the Ameucan Nuclear society.

55 Nurth Kiensnggon Aenue LiGrasrie Park. I1nots 60525

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13 ACCIDENT ANALYSES

In the other chapters of the SAR, the applicant should discuss and analyze the

safety considerations and functional requirements at a non-power Teactorihcility

for the design bases that ensure safe reactor operation and shutdown" and -

acceptable protection for the public, the operations and user sitaff and the

environment. In those chapters, the applicant should not only discuss potential

equipment malfunctions,-deviations of process variables from normal values, and

potential effects of external phenomena on the facility, but should also describe

'how'equipment will work when needed in accident situations. In Chapter 13 of the

SAR, the applicant should submit information and analyses that show that the

health and safety of the public and workers are protected and that the applicant has

considered potential radiological consequences in the event of malfunctions and

the capability of the facility'to'accommodate'such disturbances. The major

purpose ofthis chapter is for the applicant to demonstrate that the facility design

features, safety limits, limiting safety system settings, and limiting conditions for;

operation have been selected to ensure that no credible accidenticould lead to".

unacceptable radiological consequences to people or the environment.

The issue of what standards to use in evaluating accidents at'a researcwreactor

was discussed in an Atomic Safety and Licensing Appeal Board (ASLAB) decision

issued May 18, 1972; for the research reactor 'at Columbia University in New York

City. The ASLAB stated that 'as a general proposition, the Appeal Board does

not consider it desirable to use the standards of 10 CFR Part 20 for evaluating the

effects of a-p6stulated accident in-a research reactor inasmuch as they are unduly

restrictive for that purpose. The'Appeal Board strongly recommends that specific

standards for the evaluation of an accident situation in a research reactor be

formulated." The staff has not found it necessary to follow the board's

recommendation to 'develop separate criteria for evaling research reactor

accidents because'most research reactorsto date have been able to conform to the

conservative criteria of 10 CFR Part 20.

The principal safety issues that differeniiate' test reactors from 'research reaitors are

the reactor site requirements and the doses to the public'that could result from a

serious accident. For a research reactor, the results of the accident analysis have

generally been compared with 10 CFRPart 20(10 CFR 20.1 through 20.602 and

appendices for research reactors licensed before January 1, 1994, and 10 CFR

20.1001 through 20.2402 and appendices for research reactors licensed on or after

January 1, 1994). For research reactors licensed before January 1, 1994, the doses

that the staff has generally found acceptable for accident analysis results are less

than 5 rem whole bodyjand 30 rem thyroid for occupationally exposed'persons and

less than 0.5 rem whole body and 3 rem thyroid for members of the public. For

research reactors licensed on or after January 1, 1994, occupational exposure is

discussed in 10 CFR 20.1201 and public exposure is discussed in 10 CFR 20.1301.

REv. 0, 2/96 13-1 STANDARD FORMAT AND CONITENT
CHAPTER 13

In several instances, the staff has accepted very conservative accident analyses with 2

results greater than the 10 CFR Part 20 dose limits discussed above.

If the facility conforms to the definition of a test reactor, the results should be

compared with 10 CFR Part 100. As discussed in the footnotes to 10 CFR

100.11, the doses given in 10 CFR Part 100 are reference values and are not

intended to imply that the dose numbers constitute acceptable limits for emergency

doses to the public under accident conditions. Rather, they are values that can be

used for evaluating reactor sites with respect to potential reactor accidents of

exceedingly low probability of occurrence and low risk of exposure of the public to

radiation.

The accidents analyzed should range from such anticipated events as a loss of

normal electrical power to a postulated fission product release with radiological

consequences that exceed those of any accident considered to be credible. This

limiting accident is named the maximum hypothetical accident (MHA) for nonpower reactors; the details are reactor specific. Because the MHA is not expected

to occur, the scenario need not be entirely credible. The initiating event and the

scenario details need not be analyzed, but the potential consequences should be

analyzed and evaluated.

The information on credible postulated accidents should achieve the following

objectives:

* Ensure that enough events have been considered to include any accident

with significant radiological consequences. Rejection of a potential event

should be justified in the discussions.

* Categorize the initiating events and scenarios by type and likelihood of

occurrence so that only the limiting cases in each group must be

quantitatively analyzed.

* Develop and apply consistent, specific acceptance criteria for the

consequences of each postulated event.

Each postulated event should be assigned to one of the following categories, or

grouped consistently according to the type and characteristics of the particular

reactor:

* MHA

* insertion of excess reactivity (ramp, step, startup, etc.)

* loss of coolant

* loss of coolant flow

* mishandling or malfunction of fuel

_U _AR E-57 _ 13. _~.02

NUREG- 1 537, PART I 13-2 REV. 0, 2/96
ACCIDENTANALYSES

* experiment malfunction

* loss of normal electrical power

* external events

* mishandling or malfunction of equipment

The accident events in each group should be evaluated systematically to identify

the limiting event selected for detailed quantitative analysis. Limiting events in

each category should have potential consequences that exceed all others in that

group. As noted above, the MHA selected should bound all credible potential

accidents at that facility, yet should be an event that is not likely to occur during

the life of the facility.

13.1 Accident-Initiating Events and Scenarios

In this section of the SAR, the applicant 'should describe potential accidentinitiating events and scenarios for non-power reactors. For documents on general

accident scenarios and analysis, radiological consequences, and fuel types, see

Section 13.4. The following sections contain suggestions for selecting and

categorizing postulated accidents:

13.1.1 Maximum Hypothetical Accident

In general, the escape of fission products from fuel 'or fueled 'experiinents and their

release to the unrestricted environment would be the most hazardous radiological

accident conceivable at a non-power reactor. However, non-'power reactors are

designed and operated so that a fission product release is not credible for most.

Therefore, this release under accident conditions can reasonably be selected as the

MHA, which bounds all credible accidents and can'be used to illuitate the analysis

of events and consequences during the accidental release of radioactive material.

The applicant may choose to perform'sensitivity analysis of the assumptions' of the

MHA For example, reactor operating time before accident initiation may be

examined to determine the change in MHA outcome if a more realistic assumption

is made. However, these assumptions may form the basis for technical -

specification limits on the operation of the facility. The MHA could be any of the

following:. .

* A specified fraction of fuel in the core melts. (How this occurs may or may

not be specified.) - - - . . ^

* Cladding is stripped from a specified fraction of the core fuel plates or

elements.

REV. O. 2/96, -a-- - 13-3 STANDARD FORMAT AND CON TEN
CHAPTE 13

* The fuel encapsulation bursts, releasing gaseous fission products to the pool

or the air. (The failure of one fuel element in air is the MHA for a TRIGA

reactor.)

* A fueled experiment melts or fails catastrophically in the pool or in the air.

13.1.2 Insertion of Excess Reactivity

In some cases, the insertion of excess reactivity can be an initiating event that leads

to fuel or fueled experiment melting, which is the MHA. Insertion-of-excessreactivity accidents can also be used to show that limiting conditions for operation

on reactivity are justified. Some insertion-of-excess-reactivity events are the

following:

* Rapid inadvertent insertion of a portion of all excess reactivity loaded into

the reactor.

* Rapid removal of the most reactive control rod or shim rod.

* Rapid insertion of a fuel element into a vacancy in the core at the most

reactive position.

* Ramp insertion of reactivity by drive motion of the most reactive control rod

or shim rod, or ganged rods, if possible. (This event could occur during

reactor startup procedures or when the reactor is at power.)

* Failure or other malfunction of an experiment that inserts excess reactivity.

(This can be used to justify movable experiment reactivity limits.)

* Rapid increase in reactivity as a result of a change in operating parameters,

such as a surge of cold coolant

13.1.3 Loss of Coolant

In many non-power reactor designs, the loss-of-coolant accident (LOCA) is of no

consequence because decay heat in the fuel is so small as to be incapable of

causing fuel failure. In some higher power reactors (normally greater than 2 MW),

an engineered safety feature, such as an emergency core cooling system, may need

to be operable for some time after reactor shutdown to remove decay heat in the

event of a LOCA. Some initiators of LOCAs are the following:

* failure or malfunction of some component in the primary coolant loop

* failure or malfunction of an experimental facility, such as a beam tube

* failure or leak of the reactor coolant boundary

NUREG-1537. PART I 13-4 REv 0,2/96 NUREG- 1537. PART I 134 REV 0, 2/96
AC=ENTANALYSES

13.1.4 Loss of Coolant Flow

This accident is usually most limiting for forced convection-cooled non-power

reactors, where the forced flow is downward through the reactor core. -The effects

of loss of coolant flow should be considered for all non-power reactors. Upon loss

of forced downward coolant flow through the core, coolant flow in the core must

reverse to upward natural-convection cooling. During the flow reversal, heat

transfer may be inadequate in the core. Loss of coolant flow may also occur if a

foreign object obstructs a coolant flow path. Some initiators of loss of coolant

flow are the following:

* loss of electrical power -

* failure of a pump or other component in the primary coolant system

* blocking or significant decrease in flow in one or more fuel coolant channels

13.1.5 Mishandling or Malfunction of Fuel

This class of accidents represents fuel damage less severe than the MHA.

Operation with water-logged fuel is an important consideration for pulsing reactors

where the sudden addition of energy to the fuel due to a pulse may cause the water

to turn quickly to steam and damage the fuel cladding. Initiating events in this

class are the following:

* overheating of fuel during steady-power or pulsed operation

* dropping or otherwise damaging fuel in any location

* dropping, impact, or other malfunction of a non-fueled component

* operation (including pulsing) with damaged fuel, such as water-logged pin- -

or rod-type fuel

13.1.6 Experiment Malfunction ..

Tle conduct of experiments is one of the important functions of a non-power

reactor. Experiments may contain fuel, explosives, and highly reactive materials..

Failure or malfunction of experiments may initiate accidents. In some cases,

particularly for lower power non-power reactors, failure or malfunction of an

experiment may be the MHA, especially if fueled experiments are allowed by the

facility license. Initiating events for this class of accidents include the following:

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CHAvER 13

* loss of cooling capability or other malfinction in a fueled experiment

resulting in liquefaction or volatilization of the fissile component

* loss of cooling capability in a strongly absorbing non-fueled experiment

resulting in absorber failure and rapid increase in reactivity

* placement of an experiment component in an unplanned location; causing

effects that were not evaluated

* failure of an experiment containing highly reactive contents

* failure of an experiment and release of corrosive materials into the reactor

coolant

* detonation of an explosive experiment

13.1.7 Loss of Normal Electrical Power

This accident initiator could result from onsite or offsite power interruptions.

Emergency power supplies, if provided, are assumed to operate. However, the

applicant may want to analyze the effects of failure of emergency power.

13.1.8 External Events

This class of accident initiators represents some outside effect on the facility, be it

natural or caused by humans. Some initiating events in this category are the

following:

* meteorological disturbance, such as hurricane, tornado, or flood

* seismic event

* mechanical impact or collision with building

* event caused by humans, such as explosion or toxic release near the reactor

building

13.1.9 Mishandling or Malfunction of Equipment

This class of accident initiators represents failures or errors that do not fall into one

of the other categories. Some initiators in this category are the following:

* operator error at the controls

* other operator errors

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Acc'Ewr ANALYsES

*malfunction or loss of safety-related instruments or controls, such as

amnplifiers or power supplies.

'electrical fault in controlsafety rod systems

* malfunction of confinement or containment system

* rapid leak of contaminated liquid, -such as waste or primary coolant :

13.2 Accident Analysis and Determination of

Consequences

In this section of the SAR, the applicant should discuss each event giving

information consistently and systematically for gaining a clear understanding of the

specific reactor and mraking comparisons with similar reactors. Many of the steps

used to select the limiting event in each category may be semiquantitative.

However, the analyses and determination of consequences of the limiting events

should be as quantitative as possible. -The following steps are suggested:

(1) State the initial conditions of the reactor and equipment. Discuss relevant

conditions depending on fuel burnup, experiments installed,' core

configurations, or other variables. Use the most limiting conditions in the

analyses. '

(2) Identify the'causes that initiate the event; the causes may include equipment

malfunction, operator error, or a natural phenomenon or one caused by

humans. Base the scenario on a single initiating malfunction, rather than on

multiple causes.

(3) 'List the sequence of eveits,'assumed equipment operation and malfunction,

and operator actions until a final stabilized condition is ieached 'Discuss

functions and actions assumed to occur that change the course of the''

accident or mitigate the consequences, such as reactor scrams or initiation of

such engineered safety features as emergency core cooling. If credit isitaken

for mitigation of the accident conse'quences, discuss the bases used to

determine that the systems'are operable and discuss the system functions.

(4) Classify damage that might occur to components during'the'accident until

the situation is stabilized. Discuss all components and barriers that could

affect the transfer of radiation and radioactivity fromn'the reactor to the

- public and that ensure continued stability of conditions after the accident.

(5) Prepare realistic analyses to demonstrate a detailed, quantitative evaluation

of the accident evolution, including the performance of all barriers and the

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ACHAPS 13

transport of radioactive materials to the unrestricted area. Include the

assumptions, approximations, methodology, uncertainties, degree'of

conservatism, margins of safety, and both intermediate transient and ultimate

radiological conditions. Justify the methods used.' Further, make sure the

information is complete enough to allow the results to be independently

reproduced or confirmed. Demonstrate the validation of the computational

models, codes, assumptions, and approximations by comparison with

measurements and experiments when possible. Describe in detail computer

codes that are used as to the name and type of code, the way it is used, and

its validity on the basis of experiments or confirmed predictions of operating

non-power reactors. Include estimates of the accuracy of the analytical

methods. In Chapter 11, 'Radiation Protection Program and Waste

Management," of the SAR, discuss the methods and assumptions used to

analyze the release and dispersion of radioactive materials from normal

operations. Adapt those methods as appropriate for accident analyses.'

(6) Define and derive the radiation source terms, if any are involved. Include in

the source terms the quantity and type of radionuclides that could be

released, their physical and chemical forms, and the duration of potential

releases. Describe potential radiation sources that could cause direct or

scattered radiation exposure to the facility staff and the public.

(7) Evaluate the potential radiological consequences using realistic methods.

Discuss the degree of conservatism in the evaluation. For example, include

a discussion of the degree of conservatism introduced by the use of

postulated release fractions or assumption of an infinite hemispherical cloud.

Include environmental and meteorological conditions specific for the facility

site to illustrate consequences. Exposure conditions should account for the

facility staff until the situation is stabilized (including staff evacuation and

reentry), the most exposed member of the public in the unrestricted

environment until the accident conditions are terminated or the person is

moved, and the integrated exposure at the facility boundary and the nearest

permanent residence. The radiological consequences should include external

and internal exposures. Address contamination of land and water where

applicable; include exposure control measures to be initiated.

13.3 Summary and Conclusions

In this section of the SAR, the applicant should summarize the important

conclusions about the postulated accidents and the potential consequences. The

applicant should compare the projected radiological consequences with the

acceptance criteria discussed previously in this chapter The information should

demonstrate that all reasonable measures have been incorporated into the facility

NUREG- 1537, PART I 13-8 RE-V 0, 2/96
AccmDEr ANALYSES

design bases to prevent undue radiation exposures and contamination of theunrestricted environment. The discussions should show that engineered safety:-

features have been incorporated where necessary to limit consequences to

acceptable levels.

13.4 Bibliography

Non-Power Reactors

American Nuclear Society, ANS 5.1; 'Decay Heat Power in Light Water

Reactors," LaGrange Park, Illinois, 1978.

Atomic Energy Commission, "Calculation of Distance Factors for Power and Test

Reactor Sites," TID-14844, March 23, 1962.

Baker, L., Jr., and L. C. Just, Studies ofMetal-Water Reactionsat'High

Temperatures, III, 'Experimental and Theoretical Studies of the Zirconium Water

Reaction,"- ANL 6548, Argonne National Laboratory, Illinois, 1962. -

Baker, L., Jr., and R. C. Liimatakinen, "Chemical Raictionsr in Volume 2 of ATe'

Technology of Nuclear Reactor Safety, Thompson and Beckerly (eds.), The MIT

Press, Cambridge, Massachusetts, 1973, pp. 419-523.

Hunt, C. R, and DeBevee, C. J., S'Effects of Pool Reactor Accidents," General

Electric Technical Information Series, GEAP 3277, November 2, 1959.

Woodrufl W. L.,, A Kinetics and Thermal-Hydraulics Capability for the Analysis -

of Research Reactors," Nuclear Technology 64, February 1984,app. 196-206.

Radiological Consequences

International Commission on Radiological Protection, "Limits for Intakes of

Radionuclides by Workers," Publication 30, Part 1 Chapter 8, Pergamon Press,

1978/1979.

Lahti, G. P.;, et al.,!tAssessment of Gain4-Ray Exposuies DuWe to Finite Plumes,w

HealthPkysics, 41, 1981, p: 319. :

U.S. Nuclear Regulatory Commission, 'Calculation of Annual Doses to Man From

Routine Releases of Reactor Effluents for the Purpose of Eraluating Compliance

With 10 CFR Part 50, Appendix I," Regulatory Guide 1 109, October 1977.

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CHAPM 13

U.S. Nuclear Regulatory Commission, 'Atmospheric Dispersion Models for

Potential Accident Consequence Assessments at Nuclear Power Plants,"

Regulatory Guide 1.145, February 1983.

U.S. Nuclear Regulatory Commission, "Nomograms for Evaluation of Doses From

Finite Noble Gas Clouds," NUREG-0851, 1983.

Behavior of Zirconium-Hydride Fueled Reactors (TRIGA)

Atomic Energy Commission, 'In the Matter of Trustees of Columbia University in

the City of New York," Docket No. 50-208, Atomic Energy Commission Reports,

Issued May 18, 1972.

Baldwin, N. L., F. C. Foushee, and J. S. Greenwood, "Fission Product Release

From TRIGA-LEU Reactor Fuels," Seventh Biennial US. TPJGA Users

Conference, San Diego, California, 1980.

Coffer, C. O., J. R. Shoptaugh, Jr., and W. L. Whittemore, 'Stability of the UZrH TRIGA Fuel Subjected to Large Reactivity Insertion," GA-6874, General

Atomics Company, transmitted by letter dated July 25, 1967 (Docket No. 5S-163),

January 1966.

Foushee, F. C., and R. R Peters, "Summary of TRIGA Fuel Fission Product

Release Experiments," Gulf-EES-A10801, San Diego, California, September

1971.

General Atomics Company, 'Technical Foundations of TRIGA," GA-0471, San

Diego, California, August 1958. %

Kessler, W. E., et al., "Zirconium-Hydride Fuel Behavior in the SNAPTRAN

Transient Tests," Transactions of the American Nuclear Society, 9, 1966, p. 155.

Lindgren, J. R., and M. T. Simnad, "Low-Enriched TRIGA Fuel Water-Quench

Safety Tests," Transactions of the American Nuclear Society, 33, 1979, p. 276.

Shoptaugh, J. R., Jr., "Simulated Loss-of-Coolant Accident for TRIGA Reactors,"

GA-6596, Gulf General Atomic, San Diego, California, August 18, 1965.

Simnad, M. T., 'The U-ZrHl Alloy: Its Properties and Use in TRIGA Fuel,"

GA-4314, E-1 17-833, General Atomics Company, San Diego, California,

February 1980.

Sirnnad, M. T., and J. B. Dee, "Equilibrium Dissociation Pressures and

Performance of Pulsed U-ZrH Fuels at Elevated Temperature," Thermodynamics

NUREG-1537, PART I 13.10 REv 0.2/96

NUTREG- 1537, PART I 13-10 REV 0, 2/96
A~cImENT ANA.YsE-s

of Nuclear Materials, Proceedings of a Symposium, Vienna, September 4-8, 1967,

International Atomic Energy Agency, Vienna, 1968.

Simnad, M. T., F. C. Foushee, and G. B. West, 'Fuel Elements for Pulsed TRIGA

Research Reactors," Nuclear Technology, 28, 1976, pp. 31-56.

U.S. Nuclear Regulatory Commission, 'Generic Credible Accident Analysis for"

TRIGA Fueled Reactors," NUREG/CR-2387, Pacific Northwest Laboratory,

1982. -

U.S. Nuclear Regulatory Commussion, "Safety Evaluation Report on High- .

Uranium Content, Low-Enriched Uranium-Zirconium Hydride Fuels for TRIGA

Reactors," NUREG-1282, 1987.

West, G. B., et al., "Kinetic Behavior of TRIGA Reactors," GA-7882, General

Atomics Company, San Diego, California,'March 1967.

Behavior of Plate Fueled Reactors .

Bullock, J. B., 'Calculation of Maximum Fuel Cladding Temperatures for Two

Megawatt Operation of the Ford Nuclear Reactor," Memorandum Report No. 1,

Memorial Phoenix Project, Michigan, June 1962.

Forbes, S. G., et al., -Instability in the SPERT I Reactor," Preliminary Report,

IDO-16309, Idaho Operations Office, Atomic Energy Commission, Idaho Falls,

October 1956.

Knexevich, M., et al., 'Loss of Water at the Livermore Pool Type Reactor,"

Health Physics II, 1965, pp. 481-487.

Miller, R N., et al., "Report of the SPERT I Destructive Test Program on an

Aluminum, Plate-Type, Water-Moderated Reactor," IDO-16883, Idaho

Operations Office, Atomic Energy Commission, Idaho Falls, June 1964.

Nyer, W. E., et al., "Experimental Investigations of Reactor Transients,"

IDO-16285, Idaho Operations Office, Atomic Energy Commission, Idaho Falls,

April 1956. 1

Shibata, T., et al., 'Release of Fission Products From Irradiated Aluminide Fuel at

High Temperatures," Nuclear Science and Engineering, 87, 1984, pp. 405-417.

Sims, T. M., and W. H. Tabor, "Report on Fuel-Plate Melting at the Oak Ridge

Research Reactor, July 1, 1963," ORNL-TM-627, Oak Ridge National

Laboratory, Tennessee, October 1964.

REV ,2/6 1.11- ~-- SANDRD ORMA AN COTEN

REK' ,0, 2/96 . . 13-1 1 - - --- - -STAmARD FoRmAT Am CoNTEw ' I .,
CHAFF 13

U.S. Nuclear Regulatory Commission, TFuel Temperatures in an Argonaut Reactor

Core Following a Hypothetical Design Basis Accident (DBA),"

NUREG/CR-2198, Los Alamos National Laboratory, February 1981.

U.S. Nuclear Regulatory Commission, "Analysis of Credible Accidents for

Argonaut Reactors," NUREGICR-2079, Pacific Northwest Laboratories, April

1981.

U.S. Nuclear Regulatory Commission, memorandum from L. S. Rubenstein to

R. L. Tedesco, "Design Basis Event for the University of Michigan Reactor,"

June 17, 1981.

U.S. Nuclear Regulatory Commission, "Safety Evaluation Report Related to the

Evaluation of Low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use

in Non-Power Reactors," NUREG-1313, 1988.

Wett, J. F., Jr., "Surface Temperatures of Irradiated ORR Fuel Elements Cooled in

Stagnant Air," ORNL-2892, Oak Ridge National Laboratory, Tennessee, April 16,

1960.

<2

K)

NUR.EG-1537, PART! 13-12 REV 0,2196 NUREG- 1537, PART I 13-12 REV 0, 2196
14 TECHNICAL SPECIFICATIONS

In this chapter of the SAR; the applicant should discuss the development of the

facility technical specifications. This chapter of the format and content guide

discusses the contents of Chapter 14 of the SAR and presents guidance in

Appendix 14.1 on the format and content of technidal specifications for nion-power

reactors.

NRC requires each applicant for a license to operate a non-power reactor to

develop technical specifications that state the limits, operating conditions, and

other requirements imposed on facility operation to protect the environment and

the health and safety of the facility staff and the public in accordance with 10 CFR

50.36. The technical specifications'are typically derived from the facility'

descriptions and safety considerations contained in the SAR and represent a

comprehensive envelope of safe operation.

Applications for construction permits or operating licenses and renewals of

operating licenses must contain proposed technical specifications that will be

incorporated in the operating license. .During its review of the application, the

NRC staff will review the SAR and proposed technical specifications to ensure

they are complete and comprehensive and that the environment and public health

and safety will be protected. After final acceptance by the NRC staff, the technical

specifications will be included as Appendix A to the operating license.

The format and content of the technical specifications discussed in Appendix 14.1

follow the format of the 1990 revision to American National Standards

Institute/American Nuclear Society (ANSIIANS) 15.1. Examples of the format

and content of technical specifications can be found in previously accepted and

approved technical specifications for non-power reactors of similar design,

operating characteristics, site and environmental conditions, and use.

This chapter of the SAR normally is very short. The applicant should be able to

state conclusively that the technical specifications were prepared following an

accepted format, that normal operation of the reactor within the limits of the

technical specifications will not result in offsite radiation exposure in excess of 10

CFR Part 20 guidelines, and that the technical specifications limit the likelihood

and consequences of malfinctions. The reader is referred to the technical

specifications, which are in a document separate from the SAR. The technical

specifications are neither derived nor justified in this chapter of the SAR. They are

determined by the analyses that appear in the other chapters of the SAR. Each of

the technical specifications should be supported by the SAR, and it is useful to

refer to the supporting SAR analysis in the basis of each technical specification.

In Appendix 14.1, every section of ANSI/ANS 15.1 is addressed. If ANSI/ANS

15.1 should be modified or clarified to provide acceptable technical specifications,

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RE%' 0. 2196 14-1 STANDARD FORMiAT AND CoNTEN
CHAPS 14

additional guidance is given. Sections that provide acceptable guidance as written

are noted.

The guidance inthis chapter and in Appendix 14.1 pertains to technical

specifications for all types of non-power reactors. Not all of the guidance given is

applicable to a particular reactor type. The applicant should propose and justif*r

those technical specifications that are applicable to the reactor design and

utilization under consideration.

The standard format and content of technical specifications for non-power reactors

are presented in Appendix 14.1. The numbering system (Sections I through 6.8)

corresponds to the numbering system in ANSIIANS 15.1-1990.

Reference

American National Standards Institute/American Nuclear Society, ANSIIANS

IS.1, 'The Development of Technical Specifications for Research Reactors," ANS,

LaGrange Park, Illinois, 1990.

NuREc.1537, Pur I 14-2 RE". 0,2/96

NUREG-1537, PART 1 14-2 REV. O. 2196
Appendix 14.1

Format and Content of Technical Specifications

for Non-Power Reactors

STANDARD FoLmAT Am CoNITEr
Format and Content of Technical Specifications for

Non-Power Reactors

The niumbering system in this appendix (Sections 1 through'6.8) corresponds t6

the numbering system in ANSI/ANS 15.1-1990.

1 INTRODUCTION

1.1 Scope

NRC accepts the guidance provided in this section of ANSI/ANS 15.L This

section confirms that the technical specifications for non-power reactors should

include all the categories in 10 CFR 50.36 for production and utilization facilities.

1.2 Application

1.2.1 Purpose

NRC accepts the guidance provided in this section of ANSIJANS 15.1. Technical

specifications represent a set of operating requirements for a reactor that the

applicant and NRC have agreed on. 'The specifications become part of the

operating license.

1.2.2 Format

Sections of the technical specifications should be numbered as indicated in

Section 1.2.2 of ANSI/ANS 15.1.' Subsections may be left out if not applicable for

a particular reactor or may be altered if necessary, but the 'subsections included

should be arranged in consecutive numerical order.'

For individual specifications in Sections 2, 3, and 4, applicability, objective,

specification, and basis information should be included in the specified format. For

Sections 5 and 6 of the technical specifications, ANSI/ANS 15.1 -suggests that the

specifications be stated without providing applicability, objective, or basis.

Although the ANSI format is preferred, NRC will accept these sections even if:

they include applicability, objective, or basis statements. ;..,,;- - * S

Technical specifications that use the SAR as a basis'should explicitly reference the

SAR section number. _in addition, any other sources used to support the technical

specification should be explicitly referenced. ' : :'- ';

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APPENDDC 14.1

13 Definitions

NRC and the non-power reactor community have agreed on most of the definitions

given in this section of ANSI/ANS 15.1. Those applicable to a particular facility

should be included verbatim. Facility-specific definitions may be added to clarify

terms referred to in the technical specifications. Modifications and additional

definitions presented below help clarify the meaning of terms used in

ANSI/ANS 15.1.

The following definitions should be modified as indicated:

* Class A reactor operator. The term acceptable to NRC is senior reactor

operator.

* Class B reactor operator. The term acceptable to NRC is reactor

operator.

* Reactor shutdown. The reactor is shut down if it is subcritical by at least

1 dollar both in the reference core condition and for all allowed ambient

conditions with the reactivity worth of all installed experiments included.

* Reference core condition. The reference core condition is the reactivity

condition of the core when it is at 20 'C and the reactivity worth of xenon is

zero (i.e., cold, clean, and critical).

* Shutdown margin. Shutdown margin is the minimum shutdown reactivity

necessary to provide confidence that the reactor can be made subcritical by

means of the control and safety systems starting from anry permissible

operating condition. It should be assumed that the most reactive

scrammable rods and all non-scrammable rods are in their most reactive

position and that the reactor will remain subcritical without further operator

action.

Note: Shutdown margin has a single value mutually acceptable to NRC

and the applicant, to be determined on a case-by-case basis.

The following definitions should be added:

* Secured shutdown. Secured shutdown is achieved when the reactor meets

the requirements of the definition of "reactor secured" and the facility

administrative requirements for leaving the facility with no licensed reactor

operators present.

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NUREG-1537, PARTI1 2 REV. 0, 2196
TEcHNcALSPEcmcAnONS

Shutdown reactivity.' Shutdown reactivity is the value of the reactivity of

the reactor with all control rods in their least reactive positions (e.g.,

--,inserted). The value of shutdown reactivity includes the reactivity value of

all installed experiments and is determined with the reactor at ambient

conditions. -

2 SAFETY LIMITS AND LIMITING SAFETY..

SYSTEM SETTINGS

2.1 Safety Limits

All reactor licensees are required by 10 CFR 50.36(c) to specify safety limits in the

technical specifications. These safety limits should be pla6ed on important process

variables identified in the SAR as necessary to reasonably protect the integrity of

the primary barrier against the uncontrolled release of radioactivity. For nonpower reactors, the radioactivity of concern is generally'the fission products in thefuel. For heterogeneous-core non-power reactors, the primary barrier is the

cladding of fuel plates, rods, or pins. Cladding integrity could be lost by softening,

melting, blistering, or yielding to excessive internal pressure, all of which are

dependent on temperature and operating history. For homogeneous-core reactors,

this primary barrier may be the fuel matrix, the primary vessel, or some other

component that contains the fuel and the fission products.

Reactor'conditions and safety limits should be developed to avoid failure of the

fuel and should be supported by SAR analyses. Manufacturers have studied failure

modes and failure parameters during fuel, development programs. NRC has issued

staff reports in the NUREG series approving the use of some types of lowenriched uranium fuel in non-power reactors. The applicant should make ;

maximum use of the appropriate references, some of which are listed at the end of

this appendi .

The applicant should consult NUREG-1313 ifor evaluating aluminumM-clad,

aluminum matrix plate-type fuels, using both highly enriched uranium (HEU) and

new uranium-sedicide low-enriched uranium (LEU). -The report discusses

temperatuies from experimental irradiation' tests at which plate blistering has been

observed,' possible fore nnner'offailure NRC finds 530 ° an acceptable fuel

and cladding temperature limitfnot to be exceeded under any conditions of..

operation. NUREG-1313 also references tests with HEU plate fuel that have led

to similar conclusions (Beeston et aL, 1980; Gibson, 1967; Nazare et al., 1975;

Stahl, 1982). X r 

There are several reports on training reactor and isotope production, General

Atomics (TRIGA)-type fuels (NUREG-1282; Simnad 'et al., 1976 and 1981;.

REv 0,2/96 ---- -- 3 STANDARD FORmAT AND CoN1NT-
APPEmin 14 1

Simnad and West, 1986; West et al., 1986). For stainless steel-clad UZrHj 65 LEU

8.5 uranium weight percent (w/o).TRIGA fuel, stainless steel-clad UZrH,1 65 EU

(70% U-235 enriched) 8.5 w/o fuel lifetime improvement program (FLIP) TRIGA

fuel, and stainless-steel-clad UZrH1 65 LEU 20 w/o and 30 w/o TRIGA fuel,

General Atomics has shown and NRC has accepted that integrity is not

compromised under the following cases and conditions:

for cladding temperature at or less than 500 'C, peak fuel temperature at or

lessthan 1150 'C

for cladding temperature greater than 500 'C, peak fuel temperature at or

less than 950 T

For aluminum-clad UZrH1 0 LEU 8 w/o TRIGA fuel, NRC has accepted that the

peak fuel temperature should not exceed 500 'C.

For pulsed training assembled reactor (PULSTAR) types, NRC has accepted that

the U0 2 fitel temperature should not exceed 2400 'C and the Zircaloy-2 cladding

temperature should not exceed 1500 'C.

For Aerojet-General Nucleonics (AGN)-201 reactor types, NRC has accepted that

the fuel temperature should not exceed 200 0C.

The applicant should base SAR analyses on the applicable fuel developer's reported

test results to ensure fuel integrity under all operating conditions.

2.1.1 Important Process Variables

ANSI/ANS 15 1 proposes a list of parameters that may beacceptable as process

variables for non-power reactors and states that safety limits will be measurable

parameters. However, as discussed in Sections 2.1.2 and 2.1.3 (below), not all

safety limits for non-power reactors must be monitored and actually measurable.

Safety limits could be inferred from limitations on other process variables.

It is convenient in discussing fuel integrity to divide the non-power reactors into

two groups: those with engineered cooling systems (forced-convection cooling)

and those without engineered cooling systems (natural-convection cooling or no

active cooling system). Safety limits for these reactors are discussed in the next

two sections

2.1.2 Criteria-Reactors With Engineered Cooling Systems

NRC modifies this section of ANSI/ANS 15.1 as follows. Operation of the

cooling system for reactors with forced-convection cooling maintains fuel

NUREG-1537, PART I 4 Rv 0.2/96

NUREG-1537, PART I 4 REV 0, 2196
TECHNICAL SPEMciAicoN

temperature within acceptable limits to ensure cladding'integrity. Important

parameters include fuel temperature, coolant flow rate, coolant inletltemperature,

height of coolant above core, and reactor power level. These parameters should

be controlled and measured. When all values are jointly maintained within the

limits determined by the safety analyses, fuel cladding integrity will not be lost.

These parameters are important process variables on which safety limits should be

established and specified in the technical specifications. Safety limits should

preclude flow instabilities in the hottest channel and ensure that the minimum'

departure from nucleate boiling ratio (DNBR) is at least 2.0 (which has been an

acceptable margin to the onset of nucleate boiling). The analyses should range

over all physical and engineering parameters of the fuel components, the core

configurations, and the coolant systems, and should also include consideration for

uncertainties. -

For reactors that will operate with both natural-convection cooling and forcedconvection cooling, safety limits should be specified for appropriate proc'ess

variables in both modes of operation (see Section 2.1.3 below). All non-power

reactors should be designed so that both fission heat and decay heat can be

dissipated without fuel damage. The analyses also should show that the safety

limits are not exceeded during all anticipated modes of operation.'' :

2.1.3 Criteria-Reactors Without Engineered -Cooling Systems

For reactors that will be licensed to operate without forced-convection cooling,

only Section 2.1.3(2) of ANSL/ANS'15.1 is acceptable according to 10 CFR

50.36(c). NRC modifies Section 2.1.3(2) of ANSVANS 15.1 as follows. For

reactors that either circulate coolant by natural thermal convection or have no'

specific coolant or cooling systems, thermal-hydraulic coolant parameters are not

separately controllable. The applicant should ensure fuel cladding integrity and

should note appropriate parameters chosen for safety limits in the technical

specifications.

High fuel temperature is the likely precursor of fuel failure. Therefore, a maximum

allowable fuel temperature safety limit should be established below which fuel I ''

integrity is ensured. On the basis of this fuel temperature, a power level should be

calculated using an appropriate margin which ensures that the fuel remains below

the fuel temperature safety limit.*

If the license will contain a provision to measure fuel temperature, the maximum. -

fuel temperature in the core would be the parameter on which a safety limit is

established. The SAR should show the relationship between the measured fuel

temperature and the maximum fuel temperature for the proposed reactor.'

conditions. If there is no provision for measuring fuel temperature directly, the

calculated power level should be selected as the safety limit, on the basis of the

REV. 0,2/96 5. STID FORI,.L&T D Cour

REV. O. 2196 s . ST&NDAPD FoRmAT AND CoNTENT
AP'EN~iIx14 1

maximum allowable fuel temperature and appropriate margin. However, the basis

for the safety limit still should be the maximum allowable fuel temperature.

Because most TRIGA-fueled cores have at least one instrumented fuel rod, NRC

has accepted fuel temperature alone as the safety limit for these reactors. Because

the point of measured fuel temperature is normally not the point of maximum fuel

temperature, the SAR should show the relationship between the measured fuel

temperature and the maximum fuel temperature.

For plate-type fuel, the capability to measure fuel temperature is generally not

available. Therefore, the applicant should determine a fuel cladding'temperature

below which cladding damage (softening or blistering) can be precluded. The

applicant should then establish a corresponding power level, reactor conditions,

and uncertainties that limit cladding temperature below the damage limit.

For reactors without fuel elements, such as the homogeneous AGN-201 reactors,

safety limits should be based on considerations similar to'those for plate-type fuel.

The power level established in the SAR as a safety limit must provide reasonable

assurance that fission products will not be released from their confining barrier,

which could be either the fuel matrix or the fuel canister.

Safety limits should be based on the SAR. The technical specifications should

discuss the mechanism and magnitude of the fuel limitation, including a primary

reference for the fuel development studies that support the safety limit presented

(e.g., NUREG-1282 and -1313; Simnad et al., 1976). The analysis should address

authorized core configurations and limiting thermal power levels and conditions

for the reactor.

Safety limits acceptable to NRC for various reactor fuels are discussed in

Section 2.1 (above).

2.2 Limiting Safety System Settings

NRC accepts the guidance of this section of ANSI/ANS 15.1. The SAR should

address normal operating conditions, off-normal operations, and all pertinent

postulated accident scenarios. For each parameter on which a safety limit is

established by the SAR, a protective channel should be identified that prevents the

value of that parameter from exceeding the safety limit. The calculated setpoint

for this protective action, providing the minimum acceptable safety margin

considering process uncertainty, overall measurement uncertainty, and the transient

phenomena of the process instrumentation, is defined as the "limiting afety system

setting (LSSS)." Because the LSSSs are analytical limits, the protective channels

may be set to actuate at more conservative values. The more conservative values

may be established as limiting conditions for operation (LCOs).

NUREG-1537. PART I6RV.029 6 REV. O. 2/96
TECHNICAL SPECIFICATONS

Such LCOs may be determined on'the'ba'sis of experience, which has showni that

safety system channels can be set readily within 20 percent of the'n6nnal operating

value for a measured parameter, if the LSSS is noi'exceeded, without undue

interference to operations. In many cases, the LCO can even be within 10 percent

of the operating value. The SARjustification for LSSSs and LCOs should be

referenced.

2.2.1 Criteria-Reactors With Engineered Cooling Systems

NRC accepts the guidance of this section of ANSI/ANS 15.1 for the forcedconvection cooling mode of operation. For reactors licensed to operate with

forced-convection cooling, this specification should list the'LSSS derived in the

SAR for each reactor parameter for which a safety limit was established: The

bases part of this specification should indicate the SAR assumptions and limits of

uncertainty for each analyzed LSSS. For reactors licensed to operate in forcedand natural-convection cooling modes, appropriate LSSSs should be listed for

both modes.

2.2.2 Criteria-Reactors Without Engineered Cooling Systems

NRC substitutes the following guidance for Section 2.2.2 of ANSIJANS 15.1.

Section 2.1.3 (above) requires that safety limits be established by SAR Analysis for

all licensed reactors; therefore, channels should be established on the basis of SAR

analysis to not violate each of these safety limits. Calculated LSSSs defined in

Section 2.2 of ANSIANS 15.1 and in this appendix should be provided as

technical specifications. '

3 LIMITING CONDITIONS FOR OPERATIONSLCOs are derived from the safety analyses in the SAR, which provide the bases for

the LCOs. LCOs are implemented administratively or by control and monitoring.

circuitry to ensure that the reactor is not damaged, that the reactor is capable of

performing its intended function, and that no one suffers undue radiological

exposures because of reactor operations. '''

NRC accepts'the guidance of this section of ANSVANS 15.1 as amplified in the"'

sections that follow. Many'of the'LCOs have ev'olved from experience. Many are

facility specific, depending on reactor type, 'operating charactenrstics, and site location. NRC accepts the LCOs discussed in this section provided that the applicant

justifies them 'and shows the applicability'to the'specific facility'.' Additional

specifications may be appropriate for unique facility'designs or expermental

features or for additional conservatism in'operations required by the applicant or

NRC. As noted above, LCOs can in many cases be set within 10 percent of the

normal operating level of a parameter. Specifications on surveillance intervals for

REV 0.2/96 7 STANDARD FORMAT AND CONTENT

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APPENDix 14.1

LCOs and other parameters and facility design features are given in Sections 4 and

5, respectively, of ANSI/ANS 15.1 and this appendix. LCOs should be provided

as outlined in the remainder of this section.

3.1 Reactor Core Parameters

(1) Excess Reactivity

The upper limit for allowed excess reactivity should be specified. The referenced

SAR analyses should discuss all operations that require excess reactivity and the

safety implications for the excess reactivity proposed. The discussions should

include operational flexibility, potential accidents, and relationship to shutdown

margin. The SAR (Chapter 4, "Reactor Description," and Chapter 13, "Accident

Analyses") should contain a discussion of the safety implications of the excess

reactivity, including the following:

* resultant shutdown reactivity with all control rods inserted

* effects on the reactor of any credible rapid removal of a control or safety rod

* potential effects of other maximum credible rapid additions of excess

reactivity

* possible reactivity changes caused by experiment failure or displacement

* interrelationship between shutdown margin and excess reactivity

If none of the postulated events would lead to loss of fuel integritywor to

uncontrolled release of radioactivity, the proposed excess reactivity would be

acceptable.

(2) Shutdown Margin

A single value for the shutdown margin, as defined in Section 1.3 (above), should

be specified. The specification should state that compliance with the shutdown

margin takes precedence over the excess reactivity specification. In addition, other

reactor parameters that apply to the shutdown margin should be stated. These

should include cold, clean core reactivity conditions (e.g., temperature and

poisons), core configuration (e.g., fuel and control rods), and the status of

experiments (e.g., movable experiments in their most reactive state). The value

of the shutdown margin should be large enough to be readily determined

experimentally, for example, 20.5% Ak/k or Ž0.50 dollar.

<ev2

NUREG-1537, PART I 8 REv. 0.2/96

NUREG 1537,PART] 8 RFV. 0. 2196
TEcHNICAL SPEcMCATONs

(3) Pulse Limits -

Because the TRIGA design is the only pulsing reactor design NRC licenses at this

time, the specific values given in the discussion below apply only to TRIGA design

pulsing reactors. However, the general design criteria discussed below may be

applicable to other potential pulsing non-power reactor designs.

The maxinnum reactivity addition for a pulse is a license condition similar to

maximum thermal power and is determined case by case. The value should be

based on the SAR analysis for maintaining fuel integrity, which considers fuel type,

limiting core configurations, reactivity feedback coefficients, operating history,

heat capacity, and peak fuel temperature limitations. This LCO on the maximum

reactivity addition administratively gives assurance that the maximum pulse reactivity addition license condition and the safety limit on maximum fuel temperature

will not be exceeded.

The SAR should show that the maximum reactor pulse for a TRIGA reactor with

stainless steel-clad UZrHl.,S fuel would not raise the peak fuel temperature of any

element above 1000 'C (Simnad et al., 1976). (This is a conservative limit,

proposed by General Atomics and accepted by NRC, that is not to be confused

with the Safety limit temperature value.) For a TRIGA reactor with aluminum-clad

UZrH1 0 fuel, the analysis should show that the peak fuel temperature will not

exceed 500 -C for the mamum reactor pulse. The analysis should be applicable

to the siecific reactor considering its core size, operating history, fuel types,

feedback coefficients, temperature gradients and the power peaking of all

authorized core configurations. The potential effects of pulsing on in-core

experiients~or detectors should be included in the analysis. .Any required

litations on experiments should be noted in Section 3.8 ofthetechnicalspecifications.

The report by Simn ad et a]. (1981) discussiiig fuel damage at the Texas A&M

University TRIGA reactor should be reviewed to determine if reactor operating

history and power level would require a lower peak pulse fuel temperature because

of damage to the fuel during pulsing operation. -,n

There should be a limit on the total worth of the pulse rod to ensure that the worth

of the pulse rod could not exceed the maximum reactivity insertion limit and allow

an amount of reactivity to be inserted into the core that could damage the fuel.

There should be a steady-state power level above which pulses should not be

initiated. For TRIGA reactors, NRC has accepted a power level of I kW in

conjunction with an interlock that prevents movement of the steady-state control

rods when the reactor is in the pulse mode.

REv 0.2/96 9 - - -STANDARD FORMAT AND ConmT.

REV 0. 2/96 I .. - 9: -- - - - --STkNDAPmFoRmAT.,vzCoNTLN-r,
APPENDrX 14.1

For other pulsing reactors, the proposed limiting fuel temperature and reactivity<_

insertion should be justified by reference to appropriate tests and analyses.

(4) Core Configurations

The applicant should specify special core configurations, experimental facilities

internal to the core, special neutron reflectors, burnable poisons, or mixed fuel

types assumed in the SAR. The following specifications should be included in the

LCO for core configurations:

* If analysis shows the reactivity effects of waterholes in the core need to be

limited due to reactivity insertion accidents, the reactor should have a closely

packed core and acceptable vacancies in the core center and the periphery

should be described. This does not prevent the use of in-core experimental

facilities. Reactors with thermal power levels in excess of 1 MW and a

cross-sectional area of a core experimental facility greater than 16 square

inches will be licensed as testing facilities.

* No fuel should be inserted or removed from the core unless the reactor is

subcritical by more than the worth of the most reactive fuel element.

* If control rods need to be removed from the reactor core for inspection, an

LCO should state the negative reactivity necessary in the core before a

control rod can be removed.

* If analysis shows that power peaking or power density is a concern in mixed

cores, then core geometry may need to be restricted to allow certain types of

fuel assemblies or elements only in certain core positions. -

Non-power reactors should be designed with reactivity and void coefficients and a

power defect sufficiently negative that many reactor transients are inherently

counteracted to avoid loss of fuel integrity. Although the individual reactivity

coefficients and power defect are addressed in the specification below, this LCO

should be used to develop specifications on allowed core configurations to ensure

the assumptions used in the development of limits on those parameters are met.

The specified conditions of core configuration are acceptable to NRC if the SAR

shows that none of the conditions analyzed could lead to loss of fuel integrity,

uncontrolled release of radioactivity, or potential exposures exceeding

10 CFR Part 20.

NUREG-1537, PART I 10 REv. 0.2/96

NUREG- 1537, PART I 10 REV. O. 2/96
.- -..---- ITEaENICAL SPECIFCAnONS

(5) Reictivity Coefficients (Added by NRC)

No'n-power reactors should specify reactivity coefficients for fuel'temperature,

moderator temperature, and void volume and a power defect if these parameters

could vary unacceptably with reactor operation. In many cases, these

measurements need oidly'be made during the startup testing program. -The net

effect of the coefficients and the power defect'should be negative over most of the

range of reactor operations. 'The SAP. analyses of both routine operation and

poteitial accident scenanios should show that the net negative'effect of these core

characteristics is'sufficient to mitigate any anticipated event or postulated accident

scenario. Reasonable values should be designed into the reactor (e.g., by undermoderation of the neutron spectrum). Values for surveillance should be specified

for those negative reactivity coefficients and the negative power defect that can be

measured. The values of the'coefficients and the'power defecti are acceptable if

they ensure'thit the assumptions and initial conditions of the analyses are

enveloped to prevent compromise of the fuel integrity during reactor transients and

other applicable accident scenarios.

(6) Fuel Parameters (Added by NRC)

An LCO should be specified for certain fuel parameters or characteristics. . . . . ,. ............ . . , . ....... -9

Design features of the approved fuel should be included in Section 5 of the

technical specifications.- Fuel-related LCOs include the following: -

(a)' A Fuel Types'. ' --

No operation with damaged fuel except to locate sued fuel. Te

definition of damaged fuel should specify limits on longitudinal

growth, bowing, or bending,'and limits on detectable amounts of

fission products that could escape through the primaiy barrer.

'.' ' Periodic visal inspection of fuel. This specification should be clear

and explicit and should reference fuel inanufacturers' guidance or

recommendations for detecting deterioration. 'The intervals andmethods of fuel inspection should be specified in Section 4 of the

technical specifications The purpose of inspection is to detect:.

cladding deterioration that results from erosion, corrosion, or other

damage. - ' '

(b) TRIGA Fuel^ .J.';; - - ' - ,. .

- I _* ; r.;. v E!;';. .; ..

Additional technical specifications limit fuel rod elongation, bowingand

uranium burnup. Limits listed below were proposed by General Atomics

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REV. O. 2/96 11 STANDARD FoRmAT AND CoNTENT '
HsAPPENDx 14 1

and accepted by the Atomic Energy Comrnission (AEC) during initial

licensing of pulsing TRIGA reactors. If these limits are reached, the fuel

element is defined as "damaged fuel." Acceptable specifications for TRIGA

fuel for both steady-state and pulsed operation include the following:

* hjgf-For stainless steel-clad UZrH,., TRIGA fuel, the sagitta

shall not exceed 0.125 in. (0.318 cm) over the length of the cladding

in a hexagonal-grid core arrangement or 0.0625-in. (0.159-cm)

elongation over the original length of the cladding in a circular-grid

core arrangement. For aluminum-clad UZrH, 0 fuel, the limit on the

sagitta should be 0.125 in. (0.3 18 cm).

* Elongation-For stainless steel-clad UZrHL6s TRIGA fuel, the total

length of the fuel element shall not exceed its original length by more

than 0.125 in. (0.318 cm). For aluminum-clad UZrH1.

0 fuel, the limit

on elongation should not exceed 0.5 in. (1.27 cm).

* Bum M-The burnup ofuranium-235 in the UZrH fuel matrix shall

not exceed 50 percent of the initial concentration (NUREG -1282 and

Simnad and West, 1986).

(c) Materials Testing Reactor (MTR)-Type Fuel

To prevent fuel swelling there should be burnup limitations on the fuel.

Aluminum-clad aluminum-matrix MTR-type fuel plate non-power reactors

should have technical specifications that limit uranium-235 burnup or fission

density. The specifications are acceptable if they are consistent with the

SAR, which accounts for all relevant thermal-hydraulic and metallurgical

considerations. NRC is specifically concerned with the maximum burnup

limit for plate-type fuels because of the buildup of oxide on the fuel

cladding. This can be a concern when applicants apply to increase the

maximum acceptable burnup. The increased resistance to heat transfer to

the coolant may affect consequences considered in Chapter 13 of the SAR

on accident analyses. NRC has accepted uranium burnup densities in fuels

based on a uranium aluminide matrix up to a fission density of 2.3 x 10

fissions/cm3 and up to 50 percent of the initial concentration of uranium-235

(Beeston et al., 1980; Gibson, 1967; Nazare et al., 1975; Stahl, 1982).

(d) PULSTAR Fuel

Burnup of pin-type PULSTAR fuel should be limited by a specification

based on testing and the SAR analysis. NRC has accepted burnup limits up

to 20,000 MWd/ton uranium.

N1JRBG-1S37.PMT I 12 REV 0,2196

NUREG-1537,PARTIl 12 REV 0, 2/96
TECHNcAL SPEmFxcAnoNs

LCOs are acceptable if they are analyzed in the SAR and consistent with the values

given above. The analyses should verify for these fuel parameter conditions that

the fuel will not exceed safety limits for normal and off-normal operations.

3.2 Reactor Control and Safety Systems

(1) Operable Control Rods:

The number and type of operable control and 'safety rods should be specified.-No

minimum number of operable control and safety rods is prescribed for non-power.

reactors. The specification regarding the number of operable control rods is,

acceptable if the excess reactivity and shutdown margin'specifications required by

the SAR analyses can be ensured for all operating conditions. The individual or

total reactivity worths need not be specifically listed. A rod of lesser worth might

be designated the 'regulating rod" and is used as a fine power adjustment

mechanism. In some cases, the worth of a control rod(s) connected to ,an

automatic control system (which can add reactivity) may be limited toa maximumn'

amount that was assumed in the SAR in this LCO. This regulating rod need not

have scram capability, but rods without scram capability should not be iusd when:'

showing compliance with shutdown margin requirements' .(Other rods of greater

worth, with an automatic protective (scram) function, should be capable of

achieving the specified shutdown margin.

The maximum scram time should be specified for each scrammable rod. The

specification should ensure that the drop times are consistent with the SAR

analysis of reactivity required as a function of time to terminate a reactivity

addition event accounting for measurement and calculational uncertainties' 'In

most non-power reactors, for rods 2 to 3 feet long, full rod Zisertion time in the

absence of excess mechanical friction or interference is less than I second. If a

specification proposes a longer scram time, it requires'appropriate'SAR'inalysis.-

NRC finds it acceptable to shutdown a non-power reactor by intentioially

scramming the control and safety rods.

(2) Reactivity Insertion Rates

The maximum rates of adding positive reactivity should be specified for the control

and safetyrods. The specification should explicitly state that gang or multiple rod

withdrawal is allowed. Control rod(s) connected to an automatic control system

may have maximum rates of reactivity addition that differ from'the'rest'of the

control rods., The'acceptable rates should be based on the SAR, including inadvertent addition of ramp reactivity at the maximum rate for the most conservative

power, rod position, and reactor conditions.

REv. 0,2/96 - 13 STANDARD FORMAT CoN'mr

REV. 0, 2./96 - - -- -. 1'3, STANDARD FORMAT AND CON7ENT
APPENDiX 14 l

(3) Pulsed Operation

Limitations on reactivity additions are discussed in Section 3.1(3) above and need

not be repeated here. If any hardware systems require special limitations for

pulsing, they should be discussed in this section of the technical specifications.

Examples might include (a) special core configuration, (b) specific location of

pulse rod, (c) number of pulse rods, and (d) removal of in-core fueled experiments.

These specifications are acceptable if the assumptions of the SAR are ensured and

damage to the reactor by authorized pulses is precluded. -

If an experiment containing fissionable material could be damaged by reactor

pulsing, limiting specifications must be provided in Section 3.8 of the technical

specifications to preclude the event.

(4) Scram Channels

A table should specify all required scram channels and setpoints, the minimum

number of channels, other functions performed by the channeL and reactor

operating mode, such as steady-state power or pulsed, and cooling method, such

as forced- or natural-convection coolant fow. The safety limits that the scram

protects should be discussed in the basis for the table. Table 14.1 shows how the

information could be displayed. Reactor scrams should be based on the SAR.

There should be at least two completely independent power level scram channels

and they should provide diversity and redundancy.

Historically, there have been cases in which NRC has accepted power level scrams

higher than the licensed power (1.2 times licensed power level is common) if

supported by the safety analysis. This power level is the only non-power reactor

scram setpoint that, if reached, violates the license (maximum power level). Some

licensees have incorrectly interpreted this scram setpoint at higher than the licensed

power level as allowing limited operation above the license power level. Although

this operation is generally not a safety concern, the NRC staff recognizes it as a

regulatory problem. For example, if the reactor power measuring channels are out

of calibration, it is possible that the reactor has been operated at several percent

above the maximum licensed power level for a period of time. To ensure that

licensed power levels are not exceeded for non-power reactor operation, the

applicant may consider the option of having an LCO which has a power level

scram set below the licensed power level. The NRC staff has accepted another

option and, upon submittal of a license amendment request and supporting safety

analysis, has approved license amendments for non-power reactors that raise the

licensed power 10 percent above the power level at which the reactor will be

operated. The applicant can then set the reactor scram in this 10-percent power

band. This allows the operating power to remain the same while retaining the

scram setpoints within the license power limit. Safety limits and associated LSSS

NUREG-1537, PART! 14 REv 0,2/96 NUREG 1537,PARTI ZI 14 REV 0, 2196
TEacHcAL SPECMFCAMONS

- Table 14.1 Typical required scrams and power reverses*

- ' ' Mmimun

Set point** Number

Channel andFunction - Required

Periodsafety Scramifpenod s 3 sec 1

Period eVe- Rodrun inifperiod 0soc : I :

Power level safety (Imear and safety) Scram if powi> 100%/e 2

Power leve reverse (safety) Rod run in if power> 97% 1

High powe/no coolant flow Scram if flow c 56.8 1/sc (900 gpm) I

and power> 100 kW

igh power/flapper open Scram ifpower> 100kW 1

and lapper is open

Flappe closed/no coolant flow Scram if flow < 56 8 Vsec (900 gpm) I

and flapper Is closed

Software (digital) malfunction Scram upon malfunction I

Loss of high voltage to detectors Scram ifvoltage is lost

Pool water level ScramifIcvel<4.88m (16ft) . .1.

above core top

Bndge not clamped Scram when clamps are released I

Bndge radiatian level and - Scramiafradiation 2 50 mrcmnhr and I

bulding exhaust air radiation level coocntratiOn 2 2 x 010 pCihnm I

Manual scram switch Scram when switch is depressed I 1

Rod magnet power keyswitch. - - -Scrn whn magnet power is turned off ."1

Fueltnpature, Scram if teperre 2 550 °C (1022 'F) 2

Retor 66olaite dtdtemperature . Scram iftcpera c 55 C (131 -F) I

Automatic control m out of limit Rod mn m if out of specification I

ExPerinet Scram if setpoint is violated I

Loss of site power : Scram if power is lost I

,As illustrative values, the setpoints and channels listed do not apply to any one reactor . -

*Values listed are limtnmg setpolnts For operatonal conveience, setpomts may be changed to

more conservative values

based on power are'still determined by the results ofthe analysis in the SAR and

should not change. - . r . '

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REV. O. 2/96 ' ' is STANDARD FoRmAT AND CONIENT
APPENDIx 14 1

The applicant can choose to continue to have power level scram setpoints above

the licensed power level and can meet the license power level requirements by such

other means such as administrative limits.

(5) Interlocks

Required interlocks that inhibit or prevent control rod withdrawal or reactor

startup should be specified by a table (see Table 14.2 as an example). Interlocks

should be specific to the facility and should be based on the SAR. These interlocks

include the following:

* operability of area or other radiation monitors

* experimental facilities

* confinement and ventilation systems

* initial conditions for pulsing

* detected neutrons for startup

* operability of measuring channel components, such as ion chamber power

supplies and recorders as discussed in the SAR

Table 14.2 Typical required interlocks*

Minxi

Numba

Channe Required Function

Recorders not operating 3 Prevent rod withdrawal (startup inhibit)

Neutron coumt rate (startup) I Prevent rod withdrawal (startup inhibit)

if countrate 5 2 cps

Simultaneous rod withdrawal S Prevent withdrawal of 2 or more rods

Nonpulse condition I Prevent movement of pulse rod m steadystate mode

Pulse withdrawal 4 Prevent movement of standard control

rods in pulse mode

Transient withdrawal I Prevent movement of pulse rod with

reactor power above I kW

Values listed are liniting setpoints For operational convenience, setpoints may be changed to more

conservative values

NUREG-1537,PARTI 16 REV 0,2/96

V>
-._ TEcmICAL SPEccAmioNs

If the reactorwill be licensed to operate in more than one mode, the specification

should include the mode for which the interlock is required. If permanent

interlocks are established for special experiments, shields, or access control, they

should be included in the technical specifications, as described in the SAR.

(6) Backup Shutdown Mechanisms

Most non-power reactors are required to use only control and safety rods for

shutdown. If the SAR identifies a need for backup mechanisms (e.g., moderator

dump in a critical facility), they should be specified with appropriate requirements

placed on their operability (LCOs).

(7) Bypassing Channels

Any individual channels identified in items 4, 5, or 6 (above) for which bypassing is

allowed during reactor operation should be justified in the SAR and specified

under this item. -Only minimal bypassing should be permitted in safety systems' and

never in a system' that'could compromise scram capability of the other channels.

Bypassing temporary scrams or interlocks associated with experiments need not be

included in the technical specifications but should be addressed in specific

experiment protocol-

(8) Control Systems and Instrumentation Requirements for Operation (Added

by NRC)

Technical specifications for non-power reactors should have redundant and

accurate power level monitors that cover the range from subcritical source

multiplication to above the full power level. Not all monitors are required to

include scram capability (see Table 14.3 for atypical minimum set). These include

a startup channel, linear power monitor, logarithmic power monitor, and safety

channel(s). In addition, most non-power reactors have a period channel (meter),

including a period scram.. One should be specified as analyzed in the reactor

transient response section of the SAR.

Some non-power reactors with forced-convection cooling have a channel that

displays the radiation level of nitrogen-16 in the primary flow. Although the

nitrogen-16 channel is not required in the SAR for mitigating transients, it can be

*.an important channel because (a) it has greater stability than delta temperature

across the core during power changes and (b) it is not affected by changesin core

flux distribution caused by fission product buildup inthe core that can affect the

ionization chambers. If it is necessary for the operators to use the'nitrogen-16

channel in reactor operations, it should be on the'list of specified channels.

REX'. 0,2196 17 STADAJW FORMAT Cointwr

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APENDvc 14 1l

Table 14.3 Typical required minimum measuring channels*

Number

Channel Required Function

Startup 1 Monitor subcritical multiplication for tartup

Power level 2 Input for safety power level scramn

Pulse power I Input for pulse power level scram

Fuel temperar 2 Input for fuel teperature sram

Log N/period I Wide range power level and input for penod meter

and penod scram

Linear power level I Display power for control

N-16 1 Displaypowerlevel

*As illustrative values, these channels do not apply to any one reactor. Miiu channels for a

particular facility re detarmined from the SAR analysis

In past cases where digital control and safety instrumentation was used, an analog

reactor protection system was specified in the technical specifications in addition

to the digital system to provide diversity and redundancy. The technical

specifications for digital systems (including the degree of diversity and redundancy

needed) are based on the analysis in Chapter 7 of the SAR, 'Instrumentation and

Control Systems."

Specifications in this section should cover the entire channel, including readout

meters and recorders and the protective functions they perform, such as to prevent

an LSSS from being exceeded.

Each non-power reactor should have more than one power level channel indication

in the control room when operating at full power. However, because sensorsand

channel electronics might not be identical, the channels may indicate slightly

different power levels. Power level is a principal license condition, and each

applicant may consider designating a primary channel for power level monitoring.

That channel should be calibrated for thermal power in the region of maxism

licensed power and 'should be recorded in a way that allows auditing for later proof

of authorized operation within the license condition. Facility procedures should

identify this designated channel and allow for alternative designations using

analytic comparisons to achieve operational flexibility, if necessary. Technical

specifications or facility procedures that do not include this concept are acceptable

to NRC.

NUREG-1537, PART I 18 REV. 0,2/96

NUREG- 1537,PART I 18 REV. O. 2/96
TEc~mcAt SPECFICmKIIoNs

3.3 Coolant Systems

The basic systems required for cooling the fuel and other componentsfor limiting

corrosion, and for monitoring coolant radioactivity in non-power reactors should

be'specified in this section. -All non-power reactors should have the capability to

remove both fission and decay beat to ensure fuel integrity-under all potential

conditions. All reactors having forced-convection cooling systems should have

specifications ensuring operability of systems and reactor configurations for failsafe changeover from normal forced-convection to emergency forced-convection

cooling or natiral-convection cooling. An adequate heat sink, as described in the

SAR, is a necessary component of such a system.

For reactors licensed to operate in both forced- and natural-convection cooling

modes, the appropriate coolant system configurations and the relevant power

levels for both modes should be specified as analyzed in the SAR.

Not all of the following items apply to all types of non-power reactors. However,

when applicable, they should be limited by technical specifications on the basis of

the aniayses and justifications in the SAR.

(1) Shutdown Cooling or Pump Requirements

At a minimum, the'requirements for natiural-convection cooling and the operability

and status of related systems required for shutdown should be specified as LCOs.

If additional requirements are.necessary for temporary forced-convection cooling

following' reactor shutdown from extended high-power operation, the technical

specifications should state them, using the SAR as the basis.

(2) Isolation Valves

The existence, location, operability, and status of any valves required to isolate

subsystems or components for opeiational needs, including removal of decay heat,

should be specified as LCOs. ' ' -

(3) Coolant Level Limits ,

Both the coolant pressure (boiling temperature at the fuel) and adequate naturalconvection'flow depend on the level of water above the core. JIn addition, vertical.

and horizontal radiation shielding by the coolant might be necessary. Pool water

level should be in LCO for both reaisons, using the SAR as the basis.

REV 0.2/96 19 . . STANDARD FoRMAT AND CONTENT
APPENmx 14 Ii1 -

(4) Detection of Leakage or Loss of Coolant

If primary system leakage or other loss of coolant could lead to an uncontrolled

release of radioactive material (see items S and 8 below) to the environment, an

LCO should state the need for operability of leakage detection systems.

Depending on reactor design, the applicant may also want to detect loss of coolant

to prevent fuel damage. An example of this type of system would be

instrumentation that monitors the pressure difference between the primary and

secondary cooling systems at the heat exchanger to detect conditions that would

allow loss of primary coolant in the event of a heat exchanger leak. Another

example is heavy water systems that may have detectors located on pumps and

piping to detect leakage.

(5) Detection of Fission Product Activity

The technical specifications should provide for prompt detection of fission

products escaping from the fuel barrier. The methodcoduld be a radiation detector

placed in the primary coolant flow loop or a strategically located continuous air

monitor in the reactor room or in a ventilation duct. Temporary substitutions, in

case the fission product monitor is inoperable, should follow guidance in

Section 3.7.1 of ANSI/ANS 15.1. This specification may be combined with the

specification discussed in Section 3.7.1(2) on fission product monitors.

The specified fission product monitor should be able to initiate action, such as a

reactor scram, reactor room isolation, or an alarm, as appropriate. The SAR

should provide the bases and describe how fission products are distinguished from

other waterborne or airborne radioactivity.

(6) Hydrogen Concentration (Off-Gas) Limits

If the SAR has shown any of the isotopes of hydrogen (hydrogen, deuteriumni, and

tritium) to be a significant risk to personnel or the facility, an LCO should provide

for detection or adequate control, as discussed in the SAR..

(7) Emergency Core Cooling Systems

If the SAR indicates a need for supplemental core cooling to mitigate a loss-ofprimary-coolant event, the technical 'specifications should contain an LCO

requiring an operable and adequate system. The system should satis*y the cooling

requirements for the SAR scenario and should not depend on continued availability

of normal electrical service.

NUREG-1537. PART I 20 REv. 0,2/96

NUREG- 1537, PART I 20 REV. 0, 2196
TEa~cH~AL SpEcIIcAnoNS

(8) Secondary and Primary Coolant Radioactivity Limits

In addition to the prompt detection of fission products from failed fuel or

experimenttmalfunctions [see Section 3.3(5)], LCOs should limit radioactivity in

'the coolant. The technical specifications should require periodic sampling and

appropriate analyses to detect and quantify radioactivity in both the primary and

secondary coolant. The coolant should be sampled for gross activity at a short'

interval, for example, weekly, and sampled for isotope identification at a longer

interval, for example, quarterly. Trending the output of the fission product

monitor or other primary coolant continuous radiation monitor may substitute for

weekly gross activity sampling of primary coolant ifjustified by the applicant. The

purpose of this LCO is to detect deterioration of components in the primary

coolant loop, such as a control element, and leakage in a heat exchanger into the

secondary coolant loop These specifications should be stated in such a way that

significant changes in radioactivity, as defined in the SAR, trigger remedial action.

(9) Water Chemistry Requirements

To control (a) corrosion of suchcomponents as the reactor fuel, structure,-and

pool and (b) activation of impurities in the reactor coolant,'and to maintain visual

clarity of the'reactor coolant, there should be LCOs on both electrical conductivity

and pH ofthe priimay coolant. These'specifications also should apply to any

water that comes into contact with the fuel, such as water in fuel storage tanks and

pits. The explicit limits and ranges of values should be given and should be

consistent with recommended values given by both fuel vendors and in water -

chemistry guidelines The conductivity should be monitored continuously. There

should be a definite schedule for measuring pH during both operating and

shutdown periods.' The bases should clearly address the appropite ranges and

give meaningful references. The SAR should justiy the'values'for conductivity

and the pH for the particular reactor. Acceptable'ranges for these process

variables have traditionally been s5 pmhos/cm for conductivity and between 5.0

and 7.5 for pH. These values can usually be achieved by demrineralization, filtration, and good housekeeping practices, but chemical methods should be described

and specified, if applicable.

3.4 Containment or Confinement

Because accidents -that 'result in both release of steam and building overpressure

are uncommon, most non-power reactors are housed in a confinement, not a

containment. 'There should be an LCO requiring that the system specifically

described in the SAR exist as stated. The system should be operable diririg

operation and for other applicable times such as before operation and following

shutdown, as noted in Sections 3.4.1 and 3.4.2 of ANSI/ANS 15 1. If interlocks

or administrative controls to ensure operability are required, there should be

REV ,219 21 ~ ORMA ~ Cwr_

RE~V 0,2/96 .21 - ---. ' STANDARD FORMAT AND CoNTEmr
APPENDC 14 1

appropriate specifications. Whether the facility has a confinement or a

containment depends on the reactor design, operating characteristics, and facility

location. Specifications should require nominal exhaust rates for air under the

operating and accident conditions analyzed in the SAR. Specifications should limit

building leak rates to those described in the SAX

3.5 Ventilation Systems

Ventilation and exhaust flow rates and the systems to achieve the controlled

release of effluents, as analyzed in the SAR, should be specified as LCOs. These

LCOs should be established to achieve controlled release of effluents. Automatic

failsafe closure of vents should be specified for confinement systems. Provisions to

initiate controlled, filtered, and monitored exhaust and ventilation for radiological

accidents should be included. In some cases, depending on the results of the

analysis, minimum airflow rates may be LCOs.

The ventilation system should maintain a lower air pressure in the reactor room

than in adjacent spaces. Air in the reactor room should not be distributed to other

occupied spaces within buildings. The location and height of the air exhaust

system stack or release point should be specified as an LCO here or as a design

feature in Section 5 of the technical specifications. The dimensions of the stack

should be consistent with the assumptions used in the SAR to predict potential

radiation doses in the unrestricted environment. It is acceptable that the

concentration of airborne radioactivity at the point of exhaust for normal operation

be higher than the regulatory limit for restricted areas, provided that this point is

not readily accessible to the public, the analyzed doses to the public are well below

regulatory limits for unrestricted areas, and the potential doses to the facility staff

are within regulatory limits. The as low as is reasonably achievable (ALARA)

program should be applied in all analyses (see Section 3.7 below).

3.6 Emergency Power

Any requirement for emergency electrical power for non-power reactor falities

should be analyzed in the SAR on a case-by-case basis. Any necessary facility

functions, such as radiation monitoring, emergency core cooling, or isolating the

containment or the confinement, that need to be maintained if normal electrical

power is lost should be described in the SAR. If emergency power is required, an

LCO should ensure operability of the system. The technical specification should

specify automatic startup of emergency electrical power if automatic startup is

indicated in the SARI

N1JP.EG.I537, PART I 22 REv. 0,2/96 NUREG-1 537, PARTI22 REV. 0, 2/96
TEcHmCAL SpEaccxnoms

3.7 Radiation Monitoring Systems and Effluents

Monitoring systems and effluents may be addressed in the technical specifications

under separate principal headings. The following discussion is consistent with the

corresponding sections of ANSI/ANS 15.I.

3.7.1 Monitoring Systems

A separate table in the technical specifications (see Table 14.4) should list the

required radiation monitors, the function each performs (e.g., scram or

containment isolation),- the approximate location of each, the type of radiation

detected, and the alarm and/or automatic action setting, as analyzed in the SAR.

The setpoints and calibrations should be listed in terms of radiation exposure rates

and concentrations rather than as count rates that can change with calibration.

Specificicount rates for alarms and action settings can be presented in a facility

procedure that can be amended in accord with the procedures section in the

technical specifications. For specified monitors that become inoperable, the

specification should state that reactor operations may continue only if the monitor

is replaced by a substitute or portable monitor. The replacement monitor should

perform essentially the same function until the original monitor is repaired or

replaced (generally not to exceed 1 work week unless justified in the SAR). The

specification also should state that if the specified monitor was displayed in the

control room, the operator on duty should also be able to observe the temporary

monitor. The applicant should provide a table applicable to the specific facility on

the basis of the SAR.

(1) Air Monitors (Gas and Particulate)

Monitors should be'specified for both radioactive gas and those radioactive

particulates that might be airborne in the reactor room. There should be at least

one continuous air monitor (CAM) with an audible alarm and data recorders.

These monitors should be capable of alerting facility personnel to the presence of

radioactivity. They should be calibrated for anticipated radioactive species.

Potential sources of airborne radioactivity should be analyzed in the SAR.

There should be specifications requiring operability of properly calibrated effluent

monitors, preferably with recorded outputs for long-term records that provide

documentation of the concentration and total quantity of radioactive effluents, as

discussed in Section 3.7.2 below.

For reactors operated at power levels below a few hundred kilowatts, the

concentrations of airborne radionuclides may be too low to measure during normal

operation. For these, calculated concentrations of released quantities are

acceptable as specifications, using the SAR as the basis.

REV.0.296 2 STNDAR Fo~AT CoIEN

RLN,. 0. 2M 23 STANDARD FoRmAT AND CoNTENT
APPENDix 14 1

Table 14.4 Typical required radiation measuring channels*

Minimum

Number Setpoint Equal to

Channel Required Function or Less Than

Area radiation monitors 4 Alarm 0.15 mSv/br

15 Iereinir

Hot cell monitor 1 Alarm and door interlock I MSv/hr

100 mrem/hr

Reactor bridge I Alarm (isolates containment 0 5 mSvlbr

with budlding particulate) 50 mremihr

Primary coolant I Alarm 0.5 mSv/br

-50 Borehr

Buldingparticulate .I Alarm (isolates containment 2 x 109 ,uCicm3

with reactor bridge) 2-hr particulate

Building gas (Argon-41) 1 Alarm 2 x 10-5 Cikcm3

daily rese

Stack particulate I Alarm 2 x IO 4pCilcm3

2-br particulate

Stack gas (Argon-41) I Alarm 2 x 10I pCian3

daily release

4 x I pCi/cmz

annual average

*As illustrative values, these chanmels and setpotnts do not apply to any one reattor. Setpoints

for a particular facility must be determined in the SAR analysis

(2) Fission Product Monitors

The specified fission product monitor could be the CAM or the primary coolant

monitor, depending on the release scenarios analyzed in the SAR. Release of

fission products from both fuel and fueled experiments should be included. This

specification may be combined with the specification discussed in Section 3.3(5)

above.

(3) Area Monitors

There should be a specification requiring operable area monitors in and near the

reactor room. The type of radiation detected, such as gamma rays or neutrons,

should be specified. Brand names, efficiencies, and specific designs should be I>

NURE r I24-I5 7,PA R V 0219

NUREG- 1 537, PART I 24 REV 0, 2/96
TECNICAL SPEcIcAnoNs

avoided as specifications, but the range of exposure rates monitored may be

specified. These area monitors should give information on the potential exposure

rates from reactor-related radiation. Alarm and automatic action setpoints should

be specified to ensure that personnel exposures and potential doses remain Well

below limits of 10 CFR Part 20 ajnd are consistent with the facility ALARA

program.

(4) Environmental Monitors

There should be at least one environmental monitoring station near the ficility,

preferably at the site boundary or at other areas of concern, such as at population

centers or student dormitories. 'These monitors should be'specified to match the

types'of radiation anticipated and should be either in the line of sight from the air

exhaust point or downwind in the prevailing wind, as appropriate. The types of

monitors should be specified [see Section 3.7.1(4) of ANSIANS 15 1]. The

location and method of determining background readings should be discussed in

the SAR and in the basis of the specification. -The specification should state that

environmental monitors are used to verify that the potential maximum dose, anal

or other, in the unrestricted environment is within the values analyzed in the SAR.

The specification should address both potential accident scenarios and normal

operations.

3.7.2 Effluents

All radioactive species listed in Section 3.7.2 of ANSI/ANS 15.1 that are 'released

by the facility should be addressed for normal operations, and the releases should

be limited by technical specifications. NRC accepts the proposed concentration

limits, provided the SAR shows that potential doses from these cqncentrations,

comply with 10 CFR Part 20 for the maximum exposed member ofthe public on a

facility-specific basis. If the applicant proposes to limit release to 10 CFR Part 20

limits at the point of release, then the analysis of effluents in the safety analysis

report is sufficient and no technical specifications need be proposed.

Argon-41 is the principal radionuclide released by most non-power reactors. Even'

though the doses related to argon-41 are generally small, a specification should

address the average and maximum concentrations in both the restricted and unrestricted areas and the total curies (becquerels) released during a calendar year.,

The calculated potential doses to the-most exposed persons in restricted and -

unrestricted areas must conform with 10 CFR Part 20 and the facility ALARA

program. -

Because of diffusion and dispersion of the release, if the point of release is

inaccessible to the public and generally not accessed by facility staff the maximum

normal concentration at that location may be higher than the concentration allowed

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REV. 0, 2/96- 25 -STANDARD FORMAT AND CONTENT
APPENDix 14.1

in 10 CFR Part 20 for restricted areas. The SAR should show that the diffused

and dispersed release at the point of contact with members of the public is within

10 CFR Part 20 limits. The calculations in the SAR for diffusion and dispersion

should be realistic but conservative, and should be based on logical models and

specified effluent levels. Because an infinite cloud assumption is extremely

conservative for argon-41 releases, a finite cloud should be considered, as

discussed in NUREG-0851 and accepted by NRC.

3.8 Experiments

Experimental facilities should be described in the SAR, and their basic features

should be included in Section 5 ("Design Features") of the technical specifications.

The experiments to be performed in the experimental facilities need only be noted

briefly, if at all, in the SAR, unless they could present a hazard to the reactor

facility, the public, or facility staff. Any LCOs for experiments should be

performance based to ensure that no regulations are violated, that experiment

safety analysis limits are not exceeded, and that the reactor is not damaged by

experiment failure or malfunction.

Regulatory Guide (RG) 2.2, provides detailed guidance to applicants on the scope

of the discussions for experiments to be included in the SAR. The regulatory

guide also provides guidance on the technical specifications needed to govern the

experiments performed. The technical specifications should follow the guidance of

Section 3.8 of ANSI/ANS 15.1 and Section C of RG 2.2, as supplemented by the

guidance that follows.

3.8.1 Reactivity Limits

Limits should be specified on absolute values of reactivity associated with each

type of experiment: secured, unsecured, and movable (see ANSI/ANS 15.1 and

RG 2.2 for definitions). Generally, the limits on secured experiments should be

approximately twice the limits on unsecured and movable experiments, where the

latter should be no more than I dollar. The 1-dollar limit is such that inadvertent

prompt criticality is avoided even if the experiment were to fail. Movable

experiments must be clearly defined to include those to be inserted or removed

while the reactor is operating. Unsecured experiments include those installed

before reactor startup that change position or change other conditions while the

reactor continues to operate. Reactivity limits of experiments that change position

while the reactor is operating should not exceed the ability of the reactor operator

or automatic servo system to maintain control of the reactor. The specified

reactivity limits on movable experiments should not permit the violation of the

shutdown margin specification.

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NUREG-1537,PART ] 26 REV. 0, 2196
T~eaN1CAL. SPEOFICATIONS

The specified sum of the absolute values of the reactivity worths of all experiments

should not be more than twice the limit on individual secured experiments: The

value should be consistent with the SAR analysis of inadvertent reactivity

insertions, as explained in Section C.L.a. ofRG 2.2.

There should be a specification requiring that the reactor be shut down during the

changing'or moving of any secured experiment.

3.8.2 Materials

For fissile materials in experiments, limits should be specified on the allowed

thermal power and on the equilibrium or maximum inventory of specific fission

products, such as iodines and strontium. Specifications such as those indicated in

Section C.2.a of RG 2 2 are acceptable.

A specification should require double encapsulation of potentially corrosive

materials. All liquid and gas samples should be analyzed to determine if they

require double encapsulation considering such factors as (1) the effect of failure on

the reactor, (2) the radiological consequences of failure on the facility staff the

public, and the environment, and (3) the possibility that the failure would result in

an industrial hazard that could affect safe reactor operation. The failure of an

encapsulation of material that could damage the reactor should require removal

and physical inspection of potentially damaged components. it

Specifications should limit the quantity of explosive material permitted in the

experimental facilities and elsewhere in the reactor facility. -For experimental

facilities, the upper limit should be 25 mg TNT or its equivalent, as indicated in

Section C.2.d of RG 2.2. For the overall reactor facility, the upper limit should be

no higher than 100 mg TNT or its equivalent, unless a larger quantity is analyzed

in the SAR'and approved by NRC. An additional specification should require

prior testing or analyses of explosive material encapsulations to ensure no reactor.

damage in'the event of detonation, regardless of the limit..:

A specification should limit the qu antities of uinwn materiils that could

be placed in certain experimental facilities for exploratory studies. Conformance

with Section C.2.i ofRG 2.2 would be acceptable.

3.8.3 Failure and Malfunctions

Specifications that address the failure and malfunction of an experiment and limit

the experiment paramieters should be included on i case-by-case basis, as discussed

in the SAR.`-The guildahce of Section 3.8.3(2) of ANSI/ANS 15.1 should be

followed, but specifications that require compliance with regulations are redundant

and are unnecessary [see Section 3.8.3(1) of ANSI/ANS 15.1].

REV 0.2/96 27 SmMr F0IMAT ACONTENT

REV 0, 2196 -27 STANDARD FoRmAT Aw CONTENT '
APIEDix 14.1

For experiments that may off-gas, sublime, volatilize, or produce'aerosols,

standard assumptions are often specified for calculating the activity that could be

released under normal operating conditions, accident conditions in the reactor, and

accident conditions in the experiment. Such specifications ensure conservatism in

the safety analysis of the experiment. These specifications have contained such

assumptions as the following: (1) if an experiment fails and releases radioactive

gases or aerosols to the reactor bay or atmosphere, 100 percent of the radioactive

gases or aerosols escape; (2) if an effluent holdup tank isolates on a high radiation

signal, at least 10 percent of the radioactive gases or aerosols escape; (3) if the

effluent exhausts through a filter with 99-percent efficiency for 0.3-micron

particles, at least 10 percent of the vapors escape; and (4) if an experiment fails

that contains materials with a boiling point above 130 'F (54 C), the vapors of at

least 10 percent of the materials escape through an undisturbed column of water

above the core. Any particular assumptions used should be derived from the SAR.

Applicable limits for specific experiments are normally not part of the technical

specifications and should be derived from the experiment safety review discussed

in Section 6.5 below.

3.9 Facility-Specific LCOs

The LCOs discussed above apply to most non-power reactors. Each reactor may

also have technical specifications containing facility-unique LCOs. These should

be based on the SAR and facility design.

4 SURVEILLANCE REQUIREMENTS

Certain LCOs established in Section 3 of the technical specifications should be

accompanied by a'surveillance requirement in Section 4. These surveillancerelated specifications should'clearly identify the parameter or function to be

measured or tested, the method, the frequency, and the acceptable deviation or

error. Acceptable deviations could be limited by license conditions (such as

thermal power level) or by regulations (such as 10 CFR Part 20).

NRC accepts the surveillance frequencies stated in this section of ANSI/ ANS 15.1

as amplified in the following sections. The actual wording of the specifications

should not be ambiguous. Wording in ANSIIANS 15.1 has been interpreted

incorrectly by some licensees to allow the extended interval (interval not to exceed

statement) as the average. If the extended interval is used for a particular

surveillance test, a shorter interval should be used as soon afterwards as possible

to adhere to the average. The intervals of ANSIANS 15.1 should be listed in the

applicant's technical specifications.

NLJREG-1537, PART I REv 0.2196

NUREG-1537,PARTI REV 0, 219
TECHNICAL SPECICATONS

In addition to surveillance verification of LCOs, other surveillance activities should

be specified. These include such specifications as periodic pulse rod maintenance

and cleaning, thermal power level calibration, preventive maintenance and

inspection of control/safety rod drive systems, fuel element inspections, preventive

maintenance on other important components to provide assurance of operability,

and calibration of effluent monitoring systems.

If a surveillance is not required for safety while the reactor is shut down, it may be

deferred, but must be performed before reactor startup. If the reactor is not to be

operated in a particular mode (e.g., pulse mode) for an interval that exceeds the

surveillance intervals for that particular mode, surveillances not required for safety

(an example is the requirement for a standard pulse to be performed every year)

while the reactor is operated in other modes may be deferred, but must be

performed before the reactor is considered operational in the mode in which

surveillances were deferred. Scheduled surveillances that cannot be performed

while the reactor is operating may be deferred until the next planned reactor

shutdown. Surveillances that may be deferred and the reasons for deferment

should be clearly stated in the technical specifications, justified in the SAR, andnoted in the basis of the specification.

In general, any time that a reactor system or component is modified or repaired,

the surveillance for that system should be performed as part of the operability

check of the system or component. -This should be done regardless of when the

surveillance was last performed or when it is next due. This special surveillance

may change the due date of the next regularly scheduled surveillance of that type.

4.1 Reactor Core Parameters

The excess reactivity and shutdown margin LCOs specified in Section 3 of the

technical specifications are applicable for all authorized operating conditions. As

an example, for a movable experiment, the specifications for excess reactivity and

shutdown margin surveillance measurements should be based on that experiment'

being in its most reactive location In addition, other reactor parameters that affect

reactivity during operation should be explicitly specified. For the following

specified surveillance requirements, the parameters may be determined by an

appropriate combination of measurements and calculations.

(I) Excess Reactivity - ; :

Excess reactivity should be determined at least annually and ifter changes in either

the core, in-core experiments, or control rods for which the predicted change in

reactivity exceeds the absolute value of the specified shutdown margin.

REV. 0,2196 - I 29 STANDARD FoRMAT AND COETE
-

APPENDIx 14.1

(2) Shutdown Margin

The shutdown margin should be determined at least annually and after changes in

either the core, in-ore experiments, or control rods.

(3) Pulse Limits

The relationship between peak fuel temperature and inserted reactivity for pulses

should be determined when changes are made in the core (see iten 1 above).

(4) Core Configuration

Limitations on core configurations are intended to ensure that reactor physics and

thermal-hydraulic parameters specific to the core are within 'the limits analyzed in

the SAR. Core configuration parameters specified in Section 3.1(4) or in

Section 5 of the technical specifications should be met during reactor operations.

Therefore, an acceptable surveillance specification is to verify compliance with all

applicable specifications in those sections when any change occurs in the reactor

core configuration.

(5) Reactivity Coefficients

Section 3.1(5) of the technical specifications limits reactivity coefficients, which

are largely determined by reactor design and fuel type. Meaksuing and verifying

reactivity coefficients can be a difficult task. An acceptable schedule for

surveillance of reactivity coefficients is at initial reactor startup and when any

change in the reactor core configuration or fuel type requires changes in the

specifications of Section 5.

(6) Fuel Parameters

AU TRIGA fuel should be inspected for damage and all TRIGA non-instrumented

fuel should be measured for length and bend at the following frequencies:

* For non-pulsing TRIGA reactors, the fuel should be inspected and measured

on at least a 5-year cycle., Approximately 20 percent of the fuel could be

inspected and measured annually. If an annual inspection identifies damaged

fuel, then the entire core should be inspected and measured.

* For pulsing TRIGA reactors, the fuel should be inspected and measured

annually. If the reactor is pulsed infrequently (fewer than 10 pulses

annually), the annual inspection requirement may be relaxed if the relaxed

scheduled is analyzed and justified in the SAR. If the reactor is pulsed to

reactivity insertions over 4 dollars, additional inspection requirements based

NUREG-1537, PART I 30 REV 0,2/96 NUREG- 1537, PART I 30 REV 0, 2/96
. . TEaCHCAL SPECCAIONS

on the number of pulses may be necessary. Facilities in this situation should

present and justify inspection frequency requirements determined by the fuel

vendor.

Routine inspections of fuel used in AGN-201 and PULSTAR reactors have not

been required by technical specifications. If the opportunity is presented to

conduct an inspection of fuel,'such as core disassembly of an AGN-201 or

disassembly of a PULSTAR flel element to replace fuel pins, the licensee should

consider taking advantage of this opportunity. This type of inspection need not be

a technical specification requirement.' NRC may require fuel inspection as a license

condition to increase burnup limits' on'such fuels. This would be determined on a

case-by-case basis. These inspections arei qualitative, to detect evidence-of

excessive corrosion/erosion and mechanical wear or damage.

Inspections for reactors with plate fuels have not'been required by the technical

specifications except for higher power reactors that refuel frequently. However,

for reactors that remove plate fuel from service be6auie the fuel has reached its

burnup limit, there should be a requirement to inspect representative fuel elements

(e.g., 1 in every 10) for excessive corrosion/erosion, mechanical wear or damage,

or plate swelling. The surveillance procedures should follow guidance provided by

the fuel supplier, if available. In all cases, the specification should describe briefly

how the inspection will be performed.

For reactors with technical specification limits or SARI analyses imposing limits on

uranium burnup or fission density, conflrinItoxy estimates should be made at

intervals during the life of the fuel, such as at 50, 60, aid 70 percent of the fuel life

or semiannually when the NRC/Department of Energy Form 742 is submitted.

The SAR should justify the surveillance method and intervals which ensure that the

limit is not exceeded.

For reactors with HEU fuel and subject to the 10 CFR 73.6(b) exemption (selfprotection), confirmatory radiation measurements or analyses should be made at

intervals justified in the SAR.

4.2 Reactor Control and Safety Systems

(I) Reactivity Worth of Control Rods

The'integral and differential worihs of all control and safety rods 'should be

determined at initial fuel loading. Integral and differential worths should be

determined at least annually and after changes of the'co're or control rods, as noted

in Section 4.1(1) above.

REv 0,2/96 31 STANDARD FoIMAT AND CONIThT

RE%' 0, 2/96 .31 STANDARD FOMtAT ANDm NSN
APPENDDC 14.1

(2) Rod Withdrawal and Insertion Speeds

NRC accepts the guidance provided in this section of ANSI/ANS 15.1.

(3) Transient Rod and Associated Mechanism

This system should be inspected, disassembled, cleaned, and, if applicable,

lubricated annually. The reactor should be pulsed at least annually and after

changes to core or control rods, as indicated in Section 4.1(1) above, with a welldocumented reactivity insertion. If the reactor is not routinely pulsed, this

standard pulse (a reactivity addition whose results are well know) may be deferred

for more than a year, but it should be performed before resumption of normal

pulsing.

(4) Scram Times of Control and Safety Rods

A specific interval should be stated for the surveillance intervals given in

Section 4.2(4) of ANSI/ANS 15.1.

(5) Scram and Power Measuring Channels

Channel tests of all scram and power measuring channels required by technical

specifications, including scram actions with safety rod release and interlocks,

should be performed before each reactor startup following a shutdown of more

than 24 hours or following each secured shutdown. If the reactor operating

schedule calls for no secured shutdowns, the channel tests should be performed at

least quarterly. Many facilities perform these tests before each reactor startup and

NRC recommends this practice.

(6) Operability Tests

NRC accepts the guidance provided in this section of ANSI/ANS 15.1.

(7) Thermal Power Calibration for Reactors Cooled by Forced Convection

Thermal power should be calibrated at least annually; heat balance should be

verified at least monthly.

(8) Thermal Power Calibration for Reactors Not Cooled by Forced Convection

Thermal power should be calibrated at least annually. The basis should indicate

the method to be used.

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TECIRMCAL SPEcIcATIoNs

(9) Rod Inspection

The rod-drive and scram mechanisms of each control and safety rod should be

inspected annually. The poison sections of control and safety rods should be

inspected biennially for indications'of deterioration or damage. 'This can be'a:

visual inspection or an inspection that requires the rod to pass through a measuring

device which detects swelling.

4.3 Coolant Systems

Only a small fraction of the licensed non-power reactors will need to con'sider all

of the following surveillance items; The applicant should discuss the applicability

and identify the parameters that should be tested. The applicable parameters

should be specified, and the functions should be explicitly stated in the

specification.

(1) Starting Function of Enfergency Shutdown and Sump Pumps

NRC accepts the guidance provided in this section of ANSI/ANS 15.1.

(2) Test of Emergency Coolant Sources and Systems

NRC accepts the guidance provided in this'section of ANSIVANS 15.1.

(3) Inservice Inspections

If any inservice inspections of cooling system components are identified in the

SAR, they should be performed according to the manufacturers recommendations.'

If the manufacturer's recommendation is not available, the frequency should be as

established in the SAR from engineering judgment and similar component inservice

inspection requirements and experience.

(4) Analysis of Coolants for Radioactivity

Analyses for isotope identification of primary and, if applicable, secondary coolant

should be performed by sampling quarterly. Sampling weekly for gross analysis.

should be considered to establish trends to quickly identify fuel or heat exchanger

failure.

(5) Hydrogen Concentration in Off-Gas

If applicable, this test should be performed at least annuailly and after maintenance

or repair that could affect the system or measurement' instrumentation.

R~v ,219 33STANARDFORMT AD COTE_

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APPENDUC 14.1

(6) Conductivity and pH

When the reactor is operating on a routine schedule, conductivity and pH should

be measured at least weekly. This requirement could be met by a system that

monitors conductivity and pH continuously while the reactor is operating.

If the reactor is not operated for long periods, the interval between conductivity

and pH measurements may be increased to monthly if reasonable justification is

provided in the SAR.

If fuel is stored in water in separate fuel storage from the reactor pool, the pH and

conductivity of this water should be measured at regular intervals as determined

and justified in the SAR.

(7) Primary Coolant Level

If the primary coolant level above the core is not continuously displayed during

reactor operation, the priMaiy coolant level in the pool or tank should be verified

daily if the reactor is operating or before reactor startup.

(8) Primary Coolant Sensors and Channels

Channel tests of sensor operability and channels not included elsewhere in the

technical specifications that are identified in the SAR should be performed

quarterly and before startup after maintenance.

All channels should be calibrated annually and before startup after major

modification or component replacement.

4.4 Containment or Confinement

4.4.1 Containment

Few licensed non-power reactors are required to have a containment. For those

required by the SAR, the surveillance intervals given in ANSI/ANS 15.1 are

acceptable.

4.4.2 Confinement

Confinement is a system that provides a temporary holdup or controlled release of

radioactive effluents to the environment. Most non-power reactors are equipped

with confinements and should have a functional test of the overall system described

in the SAR quarterly. In addition, an efficiency test of the filters should be

NUREG-1537, PART 1 34 REv 0,2/96

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TECHNICAL SPECmFcAnoNs

performed annually or in accordance with manufacturer recommuendations and

acceptance criteria.

4.5 Ventilation Systems

Ventilation systems at most licensed non-power reactors are an integral part of the

containment or confinement systesm, and surveillance activities may be interrelated.

An operability check, including dampers and blowers, should be performed

quarterly and following repair or maintenance to declare the system operable. The

function and efficiency of filters should be tested annually or in accordance with

manufacturer's recommnendations 'and acceptance criteria and following repair or

maintenance to declare the system operable.

4.6 Emergency Electrical Power Systems

For all emergency electrical power systems, channel checks or other operability

checks should be performed before reactor startup and after maintenance.,

Maintenance should be performed according to the manufcturer's

recommendations. If the manufacturer's recommendations are not available, the

frequency should bedas determined in the SAR, .

4.6.1 Diesels and Other Devices

The shorter of thessurveillance intervals given in this section of ANSILANS 15.1 is

acceptable.

4.6.2 Emergency Batteries

The shorterof the surveillance intervals given in this section of ANSI/ANS 15.1 is

acceptable. .

4.7 Radiation Monitoring Systems and Effluents

4.7.1 Monitoring Systems

A ch nhel check should be performed daily before reactor startup. Where

physically possible, a channel test using a radiation source should be performed at

least minthly. The SAR should describe such capability.

All required radiation' m6nitoring systems, including effluent monitors, should be

calibrated at least annually according to the manufacturer's recommendations.-,

Individual systems should have separate specifications.

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REV. O. 2/96 ..35 STANDARD FoRmAT AND CoNTENT 1.,
AEpivuc 4A I

4.7.2 Effluents

Quantities of radioactive effluents released to the environment are LCOs. If the

SAR states that it is not feasible to monitor such effluents from low power reactors

in real time at the point of release, calculated releases may be substituted. The

SAR should specify surveillance methods and intervals for confirming these

releases or for verifying upper limits.

For gaseous airborne radioactive effluent, it is acceptable to confirm annual upper

limits by integrating dosimeters such as thermoluminescence dosimeters (TLDs) or

film.

For particulate airborne or waterborne radioactive effluent, it is acceptable to

confirm annual upper limits by surveillance of environmental factors given in this

section of ANSL'ANS 15.1.

4.8 Experiments

If any experiment discussed in the SAR is designed to operate with emergency

systems or with connections to the reactor protective systems, a channel check

should be specified both daily and before reactor startup when the particular

experiment is being performed. Surveillance activities for experiments that are

included in the experiment protocol and the review and approval process need not

be included explicitly in the technical specifications.

4.9 Facility-Specific Surveillance

There should be applicable surveillance specifications for any facility-specific

LCOs in Section 3.9 of the technical specifications not explicitly included in

Section 4. These surveillances should be performed to verify significant safety

features from the SAR

5 DESIGN FEATURES

The SAR forms the basis for NRC to issue an operating license for a non-power

reactor. Essential information includes the type and enrichment of fuel, core and

fuel configurations, fuel storage facilities, thermal power level, potential accident

scenarios and mitigating features, environmental conditions at the site, and other

factors. To ensure that the issued license remains valid, design features should not

be changed without prior NRC review and approval. These major design features

are noted in Section 5 of the technical specifications, if they have not already been

specified in Section 2 or 3.

_UE-57 _AR _ _6 RE 0.2

NUREG- 1 537. PART I 36 REV 0,V296
TECHNCAL SPEc'icAoNs

The NRC accepts the guidance in this section of ANSI/AS 15.1. The applicant

should provide concise but explicit information on all noted features.

6 'ADMINISTRATIVE CONTROLS

The specified information and controls on staffing and operations of the reactor

facility will ensure that the management and staff of the facility are acceptably

knowledgeable and aware of the technical requirements to operate a safe facility,.

to comply with regulations and the license conditions, and to practice a meaningful

ALARA program, which will protect the environment and the health and safety of

the public, the facility users, and the staff.

Not all owners and operators of non-power reactors will have the same

management organization or office titles. Regardless of the details of the

management organization, or of the complexity of the facility, the administrative

functions presented in this section of ANSI/ANS 15.1 should be established and

specified. The NRC accepts the ANSI/ANS 15 1 position as modified in the

sections that follow.

6.1 Organization

6.1.1 Structure

The information recommended by ANSI/ANS 15.1 should be clearly stated,

including how and when the radiation safety staff communicates'with the facility

manager and level I management to resolve safety issues.

6.1.2 Responsibility

NRC accepts the guidance provided in this section of ANSIIANS 15 1.

6.1.3 Staffing

Applicants should use the terms "reactor operator (RO)" and "senior reactor

operator (SRO)" instead of 'Class B" and "Class A," respectively (see Figure I in

ANSI/ANS 15.1).

6.1.4 Selection and Training of Personnel

Compliance with 10 CFR Part 55 is required of the licensee and licensed operators,

unless NRC has issued'an exemption ANSI/ANS 154-1988 provides additional

guidance for non-power reactors.

REV O. 2/96 37 STANDARD FORMAT AM CONTENT
APiENDDC 14 1

6.2 Review and Audit

The committee established to perform the review function may be assigned

approval authority by the facility manager or the facility manager may retain that

authority. Section 6.2 of the technical specifications should explicitly state who

holds the approval authority and should specify the committee's authority and how

it communicates and interacts with management levels 1 and 2.

6.2.1 Composition and Qualifications

One or more voting members of the committee should be from organizations other

than the one operating the reactor.

6.2.2 Charter and Rules

NRC accepts the guidance provided in this section of ANSI/ANS 15.1.

6.2.3 Review Function

The fact that this section of ANSVANS 15.1 addresses the review function

required by 10 CFR 50.59 should be explicitly stated in the technical specifications.

6.2.4 Audit Function

In addition to the emergency plan, all other required plans, such as physical

security and operator requalification, should be specified for auditing. The

requirement to audit these plans may be part of the plan itself. If that is the case,

the requirement to audit does not need to be repeated in the technial

specifications.

6.3 Radiation Safety

The technical specifications should state that 10 CFR Part 20 establishes

requirements that the radiation safety program must achieve. Additional guidance

for radiation safety programs at non-power reactors may be found in

ANS/ANS 15.11.

The authority of the radiation safety staff to interdict or terminate safety-related

activities should be stated. The technical specifications should state management's

commitment to practice an effective ALARA program. This program should apply

to facility staff, facility users, the general public, and the environment.

NUREG-1537, PART I 38 REV. 0,2/96 NUREG 1537,PARTI 38 REV. O. 2/96
TEncAL SPEczcxnoNs

6.4 Procedures

Procedures in addition to those in ANSI/ANS 15.1 should be proposed at facilities

to address operational situations recognized in the SAR. For example, if

byproduct material whose possession is authorized under the reactor license is

used in facility laboratories that are part of the reactor license and/or transferred to

other licensees, procedures for control and transfer of this byproduct material

should be part of the set of minimuc procedures required by the technical

specifications. The specifications should be written to ensure a minimum

necessary set of procedures, but should allow for future additions as necessary.

The minor modifications and temporary deviations allowed by ANSI/ANS 15.1

should not be spelled out in the technical specifications. However, the

methodology for establishing and changing procedures should be stated in the

specifications.

6.5 Experiments Review and Approval

In addition to the guidance of ANSI/ANS 15.1, the review and approval of experiments should be consistent with the guidance provided in Section C.3 ofRG 2.2

and RG 2.4. The specifications should make clear that 'established and approved

procedures' means written procedures, properly reviewed and approved. Any

changes made to these procedures should conform to Section 6.4 of

ANSI/ANS 15.1.

6.6 Required Actions

6.6.1 Action To Be Taken in Case of Safety Limit Vidlation

NRC accepts the guidance provided in this section of ANSI/ANS 15.1.

6.6.2 Action To Be Taken in the Event of an Occurrence of the

Type Identified in Sections 6.7.2(1)(b) and 6.7.2(1)(c)

The first sentence of this section of ANSIANS 15.1 states that 'reactor conditions

shall be returned to normal or the reactor shall be shut down." The specification

should be written to provide that the applicant establish, in advance, specific.,

citeria for the two alternative actions: shutdown or retun to noral. For__

example, a returm-to-noriiial event is a reactor scram resulting from a known'

cause, such as an electric transient.

REV. 0 2/96 39 . STANDARD FORMAT MD CoNrENT __
APPEDmix 14.1

6.7 Reports

6.7.1 Operating Reports

The technical specifications should state that operating reports should be sent to

the NRC Document Control Desk and that a copy should be sent to the

appropriate regional administrator.

Section 6.7.1(4) of ANSL'ANS 15.1 refers to the reporting required by 10 CFR

50.59. The specification should make reference to the rule.

6.7.2 Special Reports

The technical specifications should state that (1) special written reports of events

should be sent to the NRC Document Control Desk and that a copy should be sent

to the appropriate regional administrator and (2) special telephone reports of

events should be made to the NRC Operations Center and the regional staff.

6.8 Records

NRC accepts the guidance provided in this section of ANSI/ANS 15.1.

7 REFERENCES

American National Standards Institute/American Nuclear Society,

ANSLVANS 15.1, "The Development of Technical Specifications for Research

Reactors," ANS, LaGrange Park, Illinois, 1990.

American National Standards Istitute/American Nuclear Society,

ANSI/ANS 15.4, "Selection and Training of Personnel for Research Reactors."

ANS, LaGrange Park, Illinois, 1988.

American National Standards Institute/American Nuclear Society,

ANSLANS 15.11, "Radiological Protection at Research Reactor Facilities," ANS,

LaGrange Park, Illinois, 1987

Beeston, J. M., R R. Hobbins, G. W. Gibson, and W. C. Francis, 'Development

and Irradiation Performance of Uranium Aluminide Fuels in Test Reactors,"

Nuclear Technology 49, pp. 136-149, June 1980.

Gibson, G. W., The Development of Powdered Uranium-Aluminide Compounds

for Use as Nuclear Reactor Fuels, IN-1 133, Idaho Nuclear Corporation, Idaho

Falls, Idaho, December 1967.

NUREG-1537. PART I 40 REv 0,2/96 NUREG- 1537, PART I 40 REV 0, 2/96
TECHNICAL SPEcIFicAnoNs

Nazare, S., G. Ondracek, and F. Thummler, "Investigations on UAIx-AI

Dispersion Fuels for High-Flux Reactors," Journal of NuclearMaterials 56,

pp. 251-259, 1975.

Simnad, M. T., et al., "Fuel Elements for Pulsed TRIGA Research Reactors,"

Nuclear Technology 28, p. 31, 1976

Simnad, M. T., et al, "Interpretation of Damage to the FLIP Fuel During

Operation of the Nuclear Science Center Reactor at Texas A&M University,"

GA-Al 6613, GA Technologies, Inc., San Diego, California, December 1981.

Simnad, M. T., and G. B. West, "Postirradiation Examination and Evaluation of

TRIGA LEU Fuel Irradiated in the Oak Ridge Research Reactor," GA-Al 8599,

GA Technologies, Inc., San Diego, California, May 1986

Stahl, D., Fuelsfor Research and Test Reactors, Status Review, ANL-85-5,

Argonne National Laboratory, Argonne, Illinois, December 1982.

U.S. Nuclear Regulatory Commission, Regulatory Guide 2.2, 'Development of

Technical Specifications for Experiments in Research Reactors," November 1973.

U.S. Nuclear Regulatory Commission, Regulatory Guide 2.4, "Review of

Experiments for Research Reactors," Revision O-R, July 1976.

U.S. Nuclear Regulatory Commission, NUREG-0851, "Nomograms for

Evaluation of Doses From Finite Noble Gas Clouds," January 1983.

U.S. Nuclear Regulatory Commission, NUREG-1282, "Safety Evaluation Report

on High-Uranium Content, Low-Enriched Uranium-Zirconium Hydride Fuels for

TRIGA Reactors," August 1987.

U.S. Nuclear Regulatory Commission, NUREG-1313, 'Safety Evaluation Report

Related to the Evaluation of Low-Enriched Uranium Silicide-Aluminum

Dispersion Fuel for Use in Nonpower Reactors," July 1988.

West, G. G, M. T. Simnad, and G. L Copeland, "Final Results From TRIGA

LEU Fuel Postirradiation Examination and Evaluation Following Long-Term

Irradiation Testing in the ORR," International Meeting on Reduced Enrichment

for Research and Test Reactors, Gatlinburg, Tennessee, November 3-6, 1986,

ANLJRERTRI-TM9, pp. 259-266, Argonne, Illinois, May 1988.

kEy O.2J96 41 STAN1.1x Foivt AND CONTENr

REV 0, 2196 41 STANDARD FORMAT Am CoNTENT
15-FINANCL QUALIFICATIONS

In this chapter of the SAR, the applicant should present the financial information

submitted to NRC for a non-power reactor license to establish that the applicant is

financially qualified to own, construct, operate, and decommission a non-power

reactor. This information should be submitted along wi the application for a

construction permit and an initial operating license or along with the application

for license renewal. Financial qualifications cover three areas:

financial ability to construct the non-power reactor facility authorized by

the construction permit

* -- financial ability.to safely'operate the' facility

* . financial ability to safely decommission the facility so that NRC can

terminate the facility license at the end of the facility's use'

Financial information to be'submitted by the applicant is discussed in 10 CFR

50.33(f) and (k).- The cover letter for the construction pemnnit and operating

license application, or for the license renewal application, can refer to this chapter,

of the SAR for complete financial information. If the applicant considers its"

financial information to be proprietary, an affidavit'in accordance with 10 CFR

2.790 shall be submitted to request that the information be withheld from the

public. If possible, a non-proprietary version of the financial information should

also be submitted.

15.1 Financial Ability To Construct a Non-Power Reactor

An applicant for a construction permit to build a non-power reactor should submit

information which demonstrates that the applicant possesses, or has reasonable

assurance of obtaining, the funds necessary to cover estimated 'construction costs

and related fuel cycle costs. This informition should include estimates of

construction and fuel cycle costs'and should identify the sources of funds to'cover

these costs.

The applicant can obtain estimates of construction costs from the facility designers

as part of the design contract, from construction bids received from contractors to

build the facility, or from costs, adjusted for inflation, for similar completed

projects. However, the number of non-power reactors constructed in recent years

is limited. -Construction costs to install a non-power reactor in an existing building

should take into account the cost of the reactor and support systems and any

modifications to the existing building. The regulations in 10 CFR 50.10 allow the

construction of multipurpose buildings (e.g., construction of a college laboratory

building that will house a non-power reactor) without the issuance of a reactor

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REV. O, 2/96 15-1 STANDARD FoRmAT Am CoNTENT
CHAPTER 5

construction permit. If the building funding has been committed, the applicant

should specify the costs for that section of a multipurpose building used to house

the non-power reactor and should indicate that the costs are considered covered

costs.

The applicant can obtain fuel cycle cost estimates from analysis of the proposed

operations, as well from analysis of proposals from fuel vendors and providers of

other services needed for the fuel cycle. The applicant can also quote recent costs

of operating similar reactors. The applicant should estimate fuel cycle costs even if

the non-power reactor facility receives fuel assistance from the U S. Department of

Energy (DOE).

The applicant should discuss the sources of funds to cover these estimated costs.

If the source of funding is not committed, the applicant should discuss the

probability of acquiring the funds and the potential source(s) of the funds. The

applicant should discuss the options available to secure funding that is not

committed through the completion of the project. The applicant should provide

supporting documentation for funding that is committed. For example, if

university funds are to be used to construct the facility, the applicant could provide

a statement signed by the chief financial officer of the university. If gifts or grants

are to be used, the applicant could submit copies of these documents. This section

also applies to fuel cycle costs. If DOE will be supplying fuel for the reactor and

support for the fuel cycle, the applicant should submit a letter from DOE stating

this fact or a copy of the DOE grant that supports the fuel cycle costs. Section 1.7

of this document contains information on fuel disposal costs.

15.2 Financial Ability To Operate a Non-Power Reactor

An applicant for an operating license or for renewal of an operating license for a

non-power reactor should submit information which demonstrates that the

applicant possesses, or has reasonable assurance of obtaining, the funds necessary

to cover estimated operating costs for the duration of the license. The applicant

should provide estimates of operating costs for each of the first 5 years of

operation of the facility or for the first 5 years of the renewal period. The

applicant should also indicate the sources of funds covering these costs.

The applicant can obtain estimates of operating costs from an analysis of the

proposed operation that takes into account the operating time and the experimental

program. The applicant can exclude from the analysis those overhead services that

are provided to all departments of the university or company without internal

transfer of funding (e.g., cleaning, utilities, and in some organizations, health

physics coverage), but the applicant should indicate that these costs are excluded

and should discuss the reasons for the exclusion. The applicant should include in

the costs the overhead that is allocated to departments (e.g., a certain percentage

NUREG-1537, PART 1 15-2 REV. 0,2/96

NUREG- 1 37, PART I 15-2 REv. 0, 2196
FrNANciAL. QuALmIcATInos

of direct salaries for benefits or a percentage of the total budget). The applicant

for a new facility can use similar operating facilities to develop cost estimates.

Applicants seeking to renew licenses have actual costs from which to develop

future estimates. The 5-year estimates should be sufficiently detailed to'show

categories of spending, such as salaries, benefits and overheadiequipment, and

supplies. If possible, the'applicant should break the estimates down'by functional

area, such as reactor operations, utilization, health physics, and administration.'

The applicant should discuss the sources of funds covering these estimated costs.

If the source of the funding is not committed, the applicant should discuss the

probability of acquiring the funds and the potential source of the funds. ''The

applicant should also discuss the possibility of operating the facility without this -

funding. The applicant should provide'supporting documentation for funding that

is committed. The applicant should submit the latest financial statements of the

university or the company as part of the evidence of financial solvency and the'

ability to fund the facility.' If gifts or grants are to be used to fund operations, the

applicant should submit copies of these documents if they are available.'

A non-power reactor may be considered a commercial reactor,'-as discussed in

10 CFR 5022. If more than 50 percent ofthe annual cost of owning and

operating the facility is devoted to the production of materials, products, or energy

for sale or commercial distribution, or to the sale of services, other than research

and development or education or training, the facility would be considered a

commercial non-power reactor. Note thatfthe key is not where funding comes

from, but the cost of owning and operating the facility and the percentage of the

cost devoted to commercial activities. It is possible for a non-power reactor to be

involved in commercial activities that provide a large'portion -of the budget (e g,

90 percent), but if the cost of conducting the commercial activity is less than

50 percent ofthe cost of owning and operating the facility, the facility may be

licensed as a Class 104 facility. This arrangement allows facilities to use

commercial activities to fund research and development A commercial non-power

reactor generally would be licensed as a Class 103 facility,-in accordance with"-

10 CFR 50.22, and the licensing process would be similar to that for a power

reactor.

The NRC staff determines whether an activity is a commercial activity on a caseby-case basis. The staff has determined that some activities are commercial- such

as irradiation of gemstones to enhance color and the irradiation of silicon to'make

semiconductors; other activities may similarly be classified as commercial

In the application, the applicant should discuss any specific activities it is involved

in or plans to be involved in and should specify which are commercial activities and

which are not. The applicant should show the percentage of cost devoted to

commercial activities.

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CHAPTER 15

15.3 Financial Ability To Decommission the Facility

The information in this section applies primarily to new applications for operating

licenses. As part of the operating license application, 10 CFR 50.33(k) requires a

report indicating how reasonable assurance will be provided that funds will be

available to decommission the facility. The information to be submitted to the

NRC for decommissioning is discussed in 10 CFR 50.75(d). The decommissioning

report shall contain a cost estimate for decommissioning the facility, an indication

of which of the method or methods described in 10 CFR 50.75(e) is to be used to

provide funds for decommissioning, and how the cost estimate and funding levels

will be adjusted periodically over the life of the facility to account for changes in

the costs of such items as labor and waste disposal charges.

To identify costs and develop the cost estimate, the applicant may have to carry

out preliminary decommissioning planning to identify the amount of radioactive

waste created, the labor required to decommission the facility, and the other

supplies and costs associated with decommissioning the facility. Chapter 17,

"Decommissioning and Possession-Only Amendments," of this document contains

additional information on decommissioning planning and decommissioning plans.

Section 1.7 of this document discusses disposal of high-level radioactive waste and

spent fuel.

Acceptable methods of providing financial assurance for decommissioning of nonpower reactors are discussed in 10 CFR 50.75(e)(2).. These methods include

prepayment, an external sinking find, and a surety method, insurance, or other

guarantee method. Federal, State, or local government applicants may submit a

statement of intent containing a cost estimate for decommissioning and indicating

that funds for decommissioning will be obtained when necessary. State university

officials may use this method for non-power reactors they own. The statement of

intent must be signed by an official who has the authority to commit to spending

the necessary funds to accomplish decommissioning. It should be clearly asserted

in the statement ihat the official signing the statement has the authority to commit

to spending the funds.

The applicant can determine the estimated cost of decommissioning from an

analysis of the facility design, as well as from an analysis of estimates and actual

costs of decommissioning similar facilities. NUREG/CR-1756 contains

information on estimating the costs of decommissioning nuclear non-power

reactors.

The applicant should discuss how the cost estimate and funding levels will be

adjusted periodically over the life of the facility to account for changes in the costs

of such items as labor and waste disposal charges.. The applicant can use actual

changes in costs such as waste disposal charges or indices of costs such as changes

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FDNczaAL QuALmcAToNs

in labor and energy costs to adjust the cost estimate. This process requires that the

original cost estimate of decommissioning be in sufficient detail to allow categories

of costs that can be adjusted. Funding levels should also be adjusted. This

adjustment may require an increase in the funds needed to decommission the

fiscility and a change to the prepayment, external sinking fund, or surety method,

insurance, or other guarantee method. Applicants using a statement of intent

should ensure that the officials responsible for the statement of intent are aware

that the cost of decommissioning the facility has changed.

15.4 Reference

U.S. Nuclear Regulatory Commission, "Technology, Safety and Costs of

Decommissioning Reference Nuclear Research and Test Reactors," .UREG/CR1756, March 1982; Addendum, July 1983.

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16 OTHER LICENSE CONSIDERATIONS

In this chapter of the SAR, the applicant should discuss license considerations that

do not belong elsewhere in the SAR. One oftthese'considerations is the way'

reactor components were used in the past. -Arecent consideration discussed in this

chapter is the medical use of non-power reactors. There may be other topics that

should appear in this chapter, but'the applicant should deterrmine these on a caseby-case basis.

16.1 Prior Use of Reactor Components

This section applies primarily to applications for license renewal, since the facility

has a history of operation. However, new facilities at which components are being

used that came from other reactor facilities may have to consider prior use of

components. An example could be control rod drives from a shutdown facility that

are beingjinstalled in a new facility. Fuel provided by the Department of Energy

(DOE) for a new facility or for a facility already in operation could come from

DOE storage and have a history of prior use that must be considered Prior use

need not have taken place at the reactor for which the applicant is seeking a'

license, but the applicant should consider how the'component was used in the past.

In SAR Chapter 13,'"Accident Analyses," the applicant discusses various accident

events considered for the facility and the corponen*or systems that prevent or

limit the release of radioactive material to the facility confinement or containment

and to the environment. Th'e applicant should consider whether prior use of

components or systems could significantly degrade their capability to continue to

perform their safety functions.

Some components and systems for which prior iie' should be considered are fuel

(fuel cladding), reactivity control systems, and engineered safety features. This list

is not all inclusive, and the applicant should review the design of the facility to

ensure that all prior use of important components and systems has been

considered.

For components or systems that are identified for analysis of prior use, the

applicant should discuss the possible mechanisms that could lead to deterioration,

along with the potential effect of the mechanism on the component or system:

Some prominent deterioration mechanisms' are the following:'

* radiation

* high temperature or temperature cycling

* corrosion

* erosion

* mechanical damage

R~v.0,2/6 1-I SANDAD FRMAT/~ C~i.

RE. ok 2/96 16-1 STANDARD FopwAT AND Co-nvu
CHAP= 16

The applicant may show by analysis that the deterioration mechanisms are not

sufficient to cause damage that would interfere with the performance of safety

functions. Components should be demonstrated to be within design-life

requirements. For example, manufacturer design specifications, recommendations,

and requirements could be used to establish acceptable performance standards for

use. Analysis could show that the stress in a component caused by temperature

cycling is not sufficient to damage the component. As part of the analysis process,

the applicant may consider the performance of similar components or systems in

other non-power reactors that have longer operating histories. Component

attributes or system performance may be measured to show that deterioration

mechanisms have not damaged a component or system to a point that impairs its

safety function. Two examples are the ultrasound measurement of pipe thickness

or the measurement of control rod drop time. In some cases, periodic

measurements may be needed to determine trends in component or system

deterioration to ensure acceptable performance. The timing of periodic

measurements and component or system performance limits may form the basis of

limiting conditions for operation in the technical specifications.

The applicant should describe the regular preventive and corrective maintenance

program that provides for replacement or repair of important components or

systems as necessary. The description should include a discussion of how

components are chosen for maintenance, how maintenance intervals are chosen for

some key components, and if the manufacturer's suggestions are heeded in the

maintenance program. The success of the program in preventing malfunctions and

other failures of equipment should be discussed. If components and systems in the

maintenance program have malfunctioned or otherwise failed, the applicant should

show that these problems do not indicate that the program has broken down

16.2 Medical Use of Non-Power Reactors

The authors of the Atomic Energy Act of 1954, as amended (AEA), saw that there

might be a use for special nuclear material in medical therapy. Section 1 04a of the

AEA allows licenses to be issued for the use of utilization facilities in medical

therapy. Two regulations in 10 CFR Part 50 carry out the provisions of the AEk

10 CFR 50.21 authorizes the issuance of Class 104a licenses for medical therapy,

and 10 CFR 50.41 states that the Commission will allow the greatest amount of

therapy possible with the special nuclear material that is available. Also,

regulations in 10 CFR Part 35, Medical Use of Byproduct Material," control the

use of byproduct material introduced into the human body, such as medical

isotopes, and byproduct sources, such as teletherapy.

The use of neutron beams to treat the cancer glioblastoma multiforme and other

brain tumors was identified in the 1950s as a potential therapeutic use for nonpower reactors. This treatment is called boron neutron capture therapy (BNCT)

NtJREG-1537, PART I 16-2 REv. 0,2/96

NUREG-1537,PARTI 1 16.2 REV. O. 2/96
OTHER LICENSE CONSIDERAnON

and consists of a pre-irradiation administration of a boron compound to the patient

and the concentration of the boron compound in the tumor. 'The patient is then

exposed to a neutron beam from the non-power reactor, causing the boron to.'

fission into lithium and an alpha particle, which are heavy charged particles. In

theory, these particles cause secondary ionization that Idlls the tumor cells:

Althoxgh the regulations for medical use of byproduct material have expanded and

matured over the years with expanded use of byproduct material in medicine, there

was no development in the area of regulations for the use of special nuclear

material and non-power reactors in medical therapy. The reason for this lack is the

fact that, until the current renewed interest in BNCT, the use of non-power

reactors in the United States for medical therapy was limited to a few BNCT trials

in the late 1950s and early 1960s that did not seem to hold much promise. There

was no need to consider regulating non-power reactors for medical therapy,-until

Massachusetts Institute of Technology (MM researchers proposed almost

30 years later that new patient trials using the MIT reactor be conducted.

Improvements in the drugs that concentrate boron in tumor cells and

improvements in beam technology toward the use of epithermal beams have

convinced researchers to reconsider BNCT and the medical use of non-power

reactors.' The NRC staff considered a number of approaches to the regulation of

non-power reactors for medical use; these raiged from doing nothing (allowing'

10 CFR 50.21 and 10 CFR 50.41 to govern medical use of non-power reactors) to

conducting'full-scale rulemaldng to create a 10 CFRPart 35 equivalent in 10 CFR

Part 50. The staff decided to place'the requirements for the conduct of human

irradiations in both the medical provider's license and the non-power reactor

technical specifications.

For regulatory purposes, the NRC staff considers medical therapylat non-power

reactors to comprise two components: (1) a medical use licensee authorized to

use the neutron beam from the facility to irradiate patients and (2) the non-power

reactor that provides the neutron beanu Licensees may also have to meet Food

and Drug Administration' (DA) requirements and receive approval from FDA to -

conduct medical therapy. These requirenents are beyond the scope of this "

document.

The medical use licensee is responsible for the patient, preparation of a treatment

plan and witten directive (as defined in 10 CFR Part 35), administration of the

boron pharmaceuticals and the neutron beamn6t the patient,'supervision of the

setup and irradiation of the patient, and contiol of the byproduct material formed

in the patient's body as a result of the treatment. The regulatory requirements for

the medical use licensee are not discussed in this chapter other than to state that

the licensee's license must list the non-power reactor medical therapy treatment

facility as an additional place of use; For further details on this aspect of BNCT,

REv. 0,2/96 - 16-3 -- *--STANDARD FORMAT ND CorENr

RL%'. 0. 2/96 - . I - 16-3 - -- -- -STANDARD FoRmAT AND CoNTENT
CHATER 16

contact the Medical, Academic, and Commercial Use Safety Branch of the

Division of Industrial and Medical Nuclear Safety in the Office of Nuclear Material

Safety and Safeguards at NRC.

The non-power reactor component comprises the medical therapy treatment

facility, production of the neutron beam, physical characterization of the beam at

its interface with the patient, all health physics considerations associated with the

beam, radioactive contamination and activation of the medical therapy treatment

room and its contents, and adherence to a quality management program for the

conduct of human therapy. These are the responsibility of the non-power reactor

licensee under the reactor license. The regulatory approach for this component is

modeled after the one in 10 CFR Part 35, Subpart L concerning teletherapy.

Licensees who want to irradiate patients at non-power reactors need to receive a

Class 104a license from NRC. The application for this license should describe any

modifications to the reactor and should discuss the safety aspects of the

modifications, the design and installation of a medical therapy treatment facility,

and additions to the technical specifications and the facility license

The need for an application for a construction permit (CP) for an existing facility

depends on the amount of modification and construction required at a facility to

initiate patient irradiations. In general, if major modifications do not need to be

made to the reactor core or structure, a CP is not required However, the

applicant should recognize that failure to address safety issues at the start of a

modification could lead to delays or additional modifications at a later time.

Medical therapy treatment facilities can be installed, filters can be added to existing

beam tubes, and in most cases, additional beam tubes can be installed in the reactor

without a CP.

Modifications to the facility to conduct animal studies or tests and experiments to

develop beam characteristics do not require a Class 104a license. These changes

to the facility or procedures described in the SAR, however, must conform to the

requirements of 10 CFR 50.59. The licensee shall seek NRC approval for

modifications that require a change to the existing technical specifications.

However, there is no assurance that modifications made to the facility before

applying for a Class 104a license will be acceptable to NRC for the conduct of

patient therapy. Although not a requirement, the NRC staff suggests that final

modifications to a facility progressing from a configuration for animal studies and

beam development to human therapy be included in the application for the Class

104a license, and that these final modifications be made after the Class 104a

license is issued.

If no issues of reactor safety are involved, NRC does not review and approve the

details of beam design. Placing filters in a beam to control beam characteristics is

NUREG-1537, PART I 16-4 REv 0,2/96

NUREG- 1 37, PART I 16o4 REV 0, 2/96
OT*ER LICENSE CONSIDERATiON

allowed in most existing licenses and is not unique to medical therapy. The filter

construction and materialismust be in accordance with the license and SAR or

changes made under 10 CFR 50.59, or the licensee shall seek NRC approval of the

changes. The characteristics of the beamn necessary to conduct'treatment are

decisions that are medical in nature and are the responsibility of the 'physician and

scientists developing the beam.

In an application for a Class 104a license, the applicant should Address the

following general guidelines:

A commitment to deliver neutrons to treat patients only pursuant to a

written directive'from a physician authorized user who is specifically

authorized to perform BNCT by an NRC- or AgreemenC State-issued

medical use license.

* A'commitment to record events equivalent to "recordable events" in

10 CFR 35.2, report events equivalent to '"mis'adinistations" in 10 CFR

35.2,'and establish a written qu'ality management program for use of the

neutron beam using criteria similar to criteria specified in 10 CFR Part 35

for teletherapy.

* -Methodology to ensure that the neutron flux, fluence, and spectrum

delivered to the patient are delivered as requested by the physician

authorized user.

* Design aspects of the neutron beam delivery system that are important to

patient or user safety to prevent changes to these aspects 'without a license

amendment.

* A reactor operator and physician communication system and a method for

.terninating the treatment exposures.'

* Consideration of the anticipated activities that may alteribeain

characteristics and may require spot checks before bam use, and the spot

-.checks that will be performed in these'situations. ' '

* Interlock systems and safety precautions to prevent personnel from being

accidentally exposed to the beam in the medical therapy treatment room

and iafety precautions to be followed before, during,'and after treatment

exposures to limit occupational exposure to ionizing radiation.'

(Information on surveillances-should be included, if appropriate.)

The applicant can use a combination of design feature's and administrative

requirements and procedures for the medical therapy treatment facility to meet the

REV 0. 2/96 16-5 STANDARD FORMAT AND CONTENT
CHAPTER 16 --- -- - - -

above general guidelines. Some of the design features and administrative

requirements and procedures may become the basis for the technical specifications.

Some design features that the applicant should consider for the medical therapy

treatment facility are the following:

Personnel should be able to scram the reactor from the medical therapy

treatment facility control area, from within the medical therapy treatment

room, and by contacting the reactor control room.

* Personnel should be able to communicate from the medical therapy

treatment facility control area to the reactor control room. They should

also be able to communicate from the medical therapy treatment control

area to inside the medical therapy treatment room.

* The medical therapy treatment room shielding design shall meet the

requirements of 10 CFR Part 20 and should meet the guidelines of the

facility ALARA program for radiation exposure of persons not undergoing

treatment.

* The medical therapy treatment room should be designed so that radiation

fields inside the room from activation of the walls, floor, and ceiling and

the equipment used for treatment (e.g., treatment table and patient

supports) are held to a minimum.

* The medical therapy treatment should be designed with two independent,

redundant systems that have the capability to cut off the beam within a

short time. These systems may be shutters, each of which has the

capability to cut off the beam. Controls for the systems should be located

both outside and inside (beam cutoff only) the medical therapy treatment

room. If shutters are used, they should be designed with redundant sources

of motion (e.g., electrical and pneumatic), or each shutter should have a

different method of motion. The systems should be designed so that one of

them can be manually operated. They should be designed so that one

system will cut off the beam if the capability to operate the other system is

lost. For example, for redundant shutters, the system should be designed

so that electrical failure or low air pressure will cause the shutters to close

if they have dual sources of motion or the operable shutter will close if the

source of motion is lost to the other shutter. The systems should be

interlocked with the medical therapy treatment room entrance so that the

beam cannot be turned on unless the room is configured to prevent entry.

If the beam is on and the medical therapy treatment room is entered, the

systems should be interlocked so that this action will cause the beam to be

NtJREG-1537, PART 1 16-6 REV 0,2/96 NUREG-1537.PARTI 16-6 REV 0, 2/96
OTHe LICENSE CONSIDERATION

cut off automatically. There should be positive indication of the actuation

of the systems at the medical therapy treatment facility control area.

The medical therapy treatment room should have a shielded door, labyrinth

with unshielded door, or some other method to prevent entry during

treatment. If the door is'opened or the room is entered during treatment,

the systems used to cut off the beam should automatically operate. If the

room has a door that is motor operated, opening that door manually should

also be possible.

* If the medical therapy treatment room does'not have a method for directly

viewing the patient (such as through a lead glass window), the'medical

therapy treatment facility should be designed with redundant methods (e.g.,

multiple television cameras and monitors) for viewing the patient. A

method of emergency lighting in case of power failure should be provided.

* The medical therapy treatment room should contain a radiation monitoring

system. The primary purpose of this system is to indicate if the systems

used to cut off the beam have worked properly by ronitoring'radiation

levels in the medical therapy treatment room. This system should have

visual and audible alarms in the medical therapy treatment facility control

area and inside the medical therapy treatment room. It should be possible

to bypass these alarms during treatment to prevent continuous alarming.

The system should have a backup power supply.

* A method should be established for monitoring the beam to determine the

neutron dose being administered to the patient.

Other design approaches may also meet the'general guidelines.' The applicant

should describe its design and explain how the design conforms with the general

guidelines. -

Administrative requirements are needed along with design features to control the

conduct of medical therapy.- The following are some administrative requirements

that the applicant should consider for the conduct of BNCT and for inclusion in the

technical specifications:

* An administrative procedure should be considered that requires patients to

be referred for treatment by a written directive from a medical use licensee

authorized by.NRC or an Agreement State to use the non-power reactor

providing the irradiation.

* The responsibilities of the non-power reactor licensee and the physician

authorized user should be stated. It should be clearly stated ihat medical

REv. 0, 2/96 16-7 STANDARD FORMAT AND CONTENT

REV. O. 2/96 16-7 STAmARD Fop,iAT AND CoNmNT
CHAPSrE 16

treatment is the responsibility of the physician in charge of the therapy and <2

the medinal physicist of the medical use licensee, the medical use licensee is

responsible for the treatment plan, and the non-power reactor licensee is

responsible for delivering the radiation fluence requested in the written

directive and for providing current and accurate beam parameters to the

medical use licensee.

. Both the non-power reactor licensee and the medical use licensee shall

agree that a patient irradiation can be started. However, either the medical

use licensee or the non-power reactor licensee should have the authority to

terminate a patient irradiation.

* Procedures should provide for removing the patient from the medical

therapy treatment room in case of medical complications or equipment

failures.

* If the reactor scram in the medical therapy treatment facility control area is

not operable, an alternative method to shut down the reactor by contacting

the reactor control room may be used for a limited time to allow repair of

the medical therapy treatment control area scram circuits. The applicant

should justify this alternative method, if requested for use.

* If the usual method of indicating the status of the systems used to cut off

the beam fails, a temporary alternative method of indicating status may be

used for a limited time to allow for repairs of the primary status indication.

The applicant should justify this alternative method, if requested for use.

* The radiation monitor in the medical therapy treatment orom should be

calibrated at regular intervals as recommended by the manufaicturer or in an

accepted national standard. The system should be checked for operability

before patient irradiation. Alarm setpoints should be stated and justified.

If the normal radiation monitor fails, a temporary alternative method of

measuring radiation fields in the medical therapy treatment room may be

used for a limited time to allow for repairs of the radiation monitoring

system. The applicant should justify this alternative method, if requested

for use.

* If one of the redundant methods of viewing the patient falls during

irradiation, the irradiation in progress may continue if the medical use

licensee and the non-power reactor licensee agree.

* Limits on the maximum amount that the radiation fluence given to the

patient can exceed the fluence prescribed in the patient treatment plan

should be stated. Definitions should be developed for recordable events

NUREG-1537, PART I 16-8 RE~V 0, 2/96
OTHER LIcENsE CONSIERAnON

and misadministrations. The applicant should discuss when and how these

events will be reported to NRC.

The scrams, beam cutoff systems interlocks, status indication of beam

cutoff systems, radiation monitors, communications equipment, and any

other systems important to'the safety of the medical therapy treatment

process should be subject to periodic 'surveillance requirements if used for

therapy. Any systems undergoing maintenance should be successfully

tested for operability before patients are treated.

'A requirement for a calibration check of the beam and a fuinctional check of

beam monitors-at regular intervals should be stated. Requirements for a

calibration check of the beam or a functional check of beam- monitors or

both in the event of beam maintenance or modification should be discussed.

A requirement for calibrating the beam monitors at regular intervals should

be stated. A requirement for characterizing the beam at regular intervals

should be stated.' A calibration check of the'beam should verify that beam

intensity or neutron spectrum has not changed. Characterization of the

beam is the determination of the dose delivered by the various components

of the beam at a certain depth in the patient.' This characterization can be

done using phantoms that represent the portion of the body to be

irradiated. The phantoms are constructed of materials that interact with

radiation in a way that is similar to the way the human body interacts and

allow placement of dosimetry to determine doses. A functional check of

the beam monitors should verify that the monitors are responding to

radiation within specified accuracy. The beam monitors may be calibrated

against instruments calibrated by a secondary calibration laboratory.

* Other reactor facilities that perform measurements associated with the

conduct of medical therapy (e.g., neutron activation analysis for

determination of boron concentration in blood or tissue) should have

calibration and surveillance requirements.

* A requirement that repair, maintenance, or modification of the medical

therapy treatment facility be properly'reviewed and conducted by qualified

personnel should be stated. It should be clear what equipment (e.g.,

medical instruments, patient positioning system) is not considered part of

the facility for purposes of this requirement.

* Requirements for qualification and training of personnel to operate the

controls for the medical therapy treatment facility should be stated. The

applicant should discuss minimum instructions that must be available at the

medical therapy treatment facility control area. The applicant should

discuss the training for making changes in the operation of the medical

REV. 0. 2/96 16-9-. . STANDARD FORMAT AND CONENT
CHAv'rER. 16

therapy treatment facility that could affect the reactivity of the reactor.

NRC-licensed reactor operators and senior reactor operators should be

trained and requalified in the operation and maintenance of the facility and

associated procedures, as appropriate, including aspects of the reactoes

operation that could affect the treatment beam. The applicant should also

discuss requirements for retaining training records.

A requirement to adhere to a quality management program for the conduct

of human therapy should be stated. The requirements for the program

should parallel those of 10 CFR 35.32, 'Quality Management Program."

The program should meet the following guidelines, which are similar to the

objectives stated in 10 CFR 35.32:

- written directive before treatment is administered

- verification of the patient's identity by more than one method before

treatment

- verification that the treatment plan and related calculations agree

with the written directive

- verification that the treatment as administered agrees with the

written directive

- assurance that unintended deviations from the written directive are

identified and evaluated and that appropriate action is taken in

response to any such unintended deviation

* Written procedures required by the technical specifications should be in

place for the conduct of medical therapy treatments before starting human

irradiations.

The appendix to this chapter is a copy of the technical specifications for human

therapy at the Massachusetts Institute of Technology.

<-/

NUE-57 _AR _ 61 ~v02

NUREG- 1 537, PART I 16-10 REV 0, 2/96
Appendix 16.1

Amendment No. 27

to Facility Operating License

No. R-37

Massachusetts Institute of Technology

STANDARD FORMAT AND COmNmT
UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON. D. C. 0555:

, awFebruary 16, 1993

Docket No. 50-20

Dr. John A. Bernard

Director of Reactor Operations

Nuclear Reactor Laboratory

Massachusetts Institute of Technology

138 Albany Street

Cambridge, Massachusetts 02139 -

Dear Dr. Bernard: *

SUBJECT: ISSUANCE OF AMENDMENT NO. 27 TO FACILITY OPERATING LICENSE

NO. R-37 - MASSACHUSETTS INSTITUTE OF TECHNOLOGY RESEARCH A

REACTOR (TAC NO. M82958)

The Commission has issued the enclosed Amendment No.27 to Facility Operating

License No. R-37 for the Massachusetts Institute of Technology Research

Reactor (MITR-II). The amendment consists of changes to the Technical:

Specifications in response to your submittal dated March 10, 1992, as

supplemented on July 30, 1992, August 31, 1992, September 23, '1992,

December 22, 1992, December 30, 1992, and January 22, 1993. '

The amendment adds requirements to the Technical Specifications (TS) for the

use of the MITR-II Medical Therapy Facility beam for human therapy.

Please provide us with a final approved copy of your procedures for the

conduct of human irradiations with the MIT medical therapy facility no later

than thirty days prior to the start of human irradiations.

A copy of the related Safety Evaluation supporting Amendment No. 27 is

enclosed.

Sincerely,

bXait

Alexander Adams, Jr.( /

Senior Project Manag

Non-Power Reactors and Decommissioning Project Directorate

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Enclosures:

1. Amendment No. 27

2. Safety Evaluation

cc w/enclosures:

See next page
Massachusetts Institute of Docket No. 50-20

Technology

cc:

City Manager

City Hall

Cambridge, Massachusetts 02139

Assistant Secretary for Policy

Executive Office of Energy Resources

100 Cambridge Street, Room 1500

Boston, Massachusetts 02202

Department of Environmental

Quality Engineering

100 Cambridge Street

Boston, Massachusetts 02108

Robert G. Zamenhof, Ph.D.

Professor of Medical Physics

New England Medical Center

750 Washington Street

Box 246

Boston, Massachusetts 02111
UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, 0 C 20555

MASSACHUSETTS INSTITUTE OF TECHNOLOGY' '

DOCKET NO.'50-20

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 27

License No. R-37

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. tThe'application for'amendment to Facility Operating License No. R-37

filed by the Massachiusetts Institute of Technology (the licensee),

''dated March 10,'1992, as supplemented on July 30, 1992, August 31,

*1992, September 23,1992, December 22, 1992, December 30, 1992, and

January 22, 1993, complies with the standards and requirements of the

Atomic Energy Act of 1954, as amended (the-Act), and the Commission's

rules and regulations as set forth in 10 CFR Chapter I;'

B. The facility will operate in conformity with the application, the

provisions of the Act, and the rules and regulations of the

Commission;

C. There is reasonable assurance: (i) that the activities authorized by

this amendment can be conducted without endangering the health and

safety of the public, and (ii) that such activities will be conducted

in compliance with the Commission's regulations4

D. The issuance of this amendment will not be inimical to the common

defense and security or to the health and safety of the public;

E. The issuance of this amendment is in accordance with 10 CFR Part 51

of the Commission's regulations and all applicable requirements have

been satisfied; and

F. Prior notice of this amendment was not required by 10 CFR 2.105(a)(4)

and publication of notice for this amendment is not required by

10 CFR 2.106(a)(2).
- 2 -

2. Accordingly, the license is amended by changes to the Technical

Specifications as indicated in the enclosure to this license amendment,

and paragraph 2.C.(2) of Facility Operating License No. R-37 is hereby

amended to read as follows:

(2) Technical Specifications

The Technical Specifications contained in Appendix A, as revised

through Amendment No-. 27, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the

Technical Specifications.

3. This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Seymour H. Weiss, Director

Non-Power Reactors and Decommissioning

Project Directorate

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Enclosure:

Appendix A Technical

Specifications Changes

Date of Issuance: February 16, 1993

k\-i-
ENCLOSURE TO LICENSE AMENDMENT NO. 27

FACILITY OPERATING LICENSE NO. R-37

DOCKET NO. 50-20

Replace the following pages of the Appendix A Technical Specifications with

the enclosed pages. The revised pages are identified by Amendment number and

contain vertical lines indicating the areas of change.

Remove Pages Insert Pages

iii iii

6-21

to

6-33

7-32 7-32
6. EXPERIMENTS

6.1 General Experiment Criteria ............................................ 6-1

6.2 In-Core Cryostat ............................................ 6-8

6.3 Sodium Subassembly for Fast Neutron Spectrum Facility .............. 6-14

6.4 Closed-Loop Control Systems ............................................ 6-17

6.5 Generation of Medical Therapy Facility Beam for Human Therapy ........ 6-21

7. ADMNSTRATIVE CONTROLS

7.1 Responsibility ............................................ 7-1

7.2 Reactor Staff Organization .................... ........................ 7-4

7.3 Reactor Staff Qualifications....................................................... 7-6

7.4 Retraining and Replacement Training ......................................... 7-8

7.5 Review ... 7-9

7.6 Action to be Taken in the Event of an Abnormal Occurrence ................. 7 -18

7.7 Action to be Taken if a Safety Limit is Exceeded ........................ 7-19

7.8 Operating Procedures.............................................................7-20

7.9 Experiment Approval Procedures ....................................... 7-22

7.10 Radiation Protection Prgram . 7-24

7.11 Security Program ................... 7-25

7.12 Records Retention .................... ; .7-26

7.13 Plant Reporting Requirements ................... 7-28

Amendment No. 27

111
6.5 Generation of Medical Therapy Facility Beam for Human Therap2

Applicabilit

This specification applies solely to the generation of the medical therapy facility beam

for the treatment of human patients. It does not apply to any other use of the medical therapy

facility and/or its beamL Surveillances listed in this specification are only required if human

therapy is planned for the interval of the surveillance. However, in the event of a hiatus in

the scheduled performance of any given surveillance, that surveillance shall be performed

prior to the initiation of human therapy during the interval in question.

Obje~ctv. -- - -

To provide for the protection of the public health and safety by ensuring that patients

arm treated in accordance with the treatment plan established by the physician authorized user

and that the ALARA principle is observed for all non-therapeutic radiation exposures.

I. Patients accepted for treatment shall have been referred by written directive from a

physician authorized user of NRC Medical Use Licensee No. 20-03857-06 or of any

other medical use licensee that has been similarly authorized by NRC to utilize the

MIT Research Reactor's Medical Therapy Facility beam for neutron capture therapy

for humans.

2. All medical treatments, including irradiations and analyses of the neutron capture

agents in the patients, are the responsibility of the physician authonzed user in charge

of the therapy and the medical physicists from the NRC-licensed medical center. The

Massachusetts Institute of Technology is only responsible for providing current and

accurate beam charactensuc parameters to the medical use licensee and for delivery of

the desired radiation fluence as requested in the written directive. Before the start of

a therapy, both the certified medical physicist and the Director of the Nuclear Reactor

Laboratory, or his designate, must agree that the therapy can be initiated. The

physician authorized user is responsible for monitoring the therapy and for directing

its termination. However, a radiation therapy can also be terminated at any time if

Amendment No. 27

6-21
either the physician authonzed user or the NRL Director, or their designates, judge

that the therapy should be terminated.

3. It shall be possible to initiate a minor scram of the reactor from a control panel located

in the medical therapy facility area. In the event that the medical facility minor scram

is inoperable, it shall be acceptable to use one of the control room scrams via

communication with the reactor operator as a temporary means of satisfying this

provision. Use of this temporary provision is limited to seven consecutive working

days.

4. Access to the medical therapy facility shall be controlled by means of the shield door

located at its entrance.

5. The following features and/or interlocks shall be operable:

(a) An interlock shall prevent opening of the shutters that control beam delivery

unless the medical therapy facility's shield door is closed.

(b) The shutters that control beam delivery shall be interlocked to close

automatically upon opening of the medical therapy facility's shield door.

(c) The shutters that control beam delivery shall be designed to close

automatically upon failure of either electric power or on low air pressure if the

shutter is operated pneumatically.

(d) Shutters that control beam delivery and that are normally pneumaticallyoperated shall, in addition, be designed for manual clostre.

(e) It shall be possible to close the shutters that control beam delivery from

within the medical therapy facility.

6. Each of the shutters that controls beam delivery shall be equipped with a light that

indicates the status of the shutter. These lights shall be visible at the medical therapy

facility's local control panel. In the event of a status light malfunction, it shall be

acceptable to use the affected shutter provided that an alternate means of verifying

position is available. Use of this alternate means of shutter position verification is

limited to seven consecutive working days.

Amendment No. 27

6-22

K'
7. The medical therapy facility shall be equipped with a monitor that provides a visual

indication of the radiaton level within the facility, that indicates both within the

facility and at the local control panel, and that provides an audible alarm both within

the facility and at the local control panel.

(a) This radiation monitor shall be equipped with a backup power supply such as

the reactor emergency power system or a battery.

(b) This radiation monitor shall be checked for proper operation by means of a

check source on the calendar day of and prior to any patient irradiation.

(c) This radiation monitor shall be calibrated quarterly.

(d) The audible alarm shall be set at or below 50 mR/hr. This monitor andlor its

alarm may be disabled once the medical therapy room has been searched and

secured, such as is done immediately prior to initiation of patient therapy. If

this is done, the monitor and/or its alarm shall be interlocked so that they

become functional upon opening of the medical therapy facility's shield door.

(e) In the event that this monitor is inoperable, personnel entering the medical

therapy facility shall use either portable survey instruments or audible alarm

personal dosimeters as a temporary means of satisfying this provision. These

instruments/dosimeters shall be in calibration as defined by the MIT Research

Reactor's radiation protection program and shall be source-checked daily

prior to use on any day that they are used to satisfy this provision. Use of

these instruments/dosimeters as a temporary means of satisfying this

provision is limited to seven consecutive working days.

8. An intercom or other means of two-way communication shall be operable both

between the medical therapy facility control panel and the reactor control room, and

also between the medical therapy facility control panel and the interior of the facility.

The latter is for the monitoring of patients.

9. It shall be possible for personnel monitoring a patient to open the medical therapy

facility's shield door manually.

Amendrment No. 27

6-23
10. It shall be possible to observe the patient through both a viewing port and by means

of a closed-circuit TV camera. Both methods of patient visualization shall be

operable at the outset of any patient irradiation. Should either fail during the

irradiation, the treatment may be continued at the discretion of the physician

authorized user. Adequate lighting to permit such viewing shall be assured by the

provision of emergency lighting.

11. The total radiation fluence delivered by the medical therapy facility beam as measured

by on-line beam monitors shall not exceed that prescribed in the patient treatellCit plan

by more than 20%. The Latment is normally delivered in fractions in accordance

with standard practice for human therapy. The 20% criterion applies to the sum of

the radiation fluences associated with all fractions in a given treatment plan. A

criterion of 30% applies to the difference between the administered and prescribed

fluence for any given week (seven consecutive days). Finally. if the treatment

consists of three or fewer fractions, then a criterion of 10% shall apply.

12. The following interlocks or channels shall be tested at least monthly and prior to

treatment of human patients if the interlock or channel has been repaired or

deenergized:

Interlock or Channel Sureillane

a) Medical therapy facility minor scram Scram test

b) Shutters will not open unless Operational test

shield door is closed

c) Shutters close upon both manual and Operational test

automatic opening of shield door

d) Shutters close on loss of electical Operational test

power and reduction of pressure

in pneumatic operators, if applicable

e) Manual closure of pneumatic shutters Operational test

Amendment No. 27

6-24

KI
f) Shutters can be closed manually Operational test

'from withi the facility

g) Shutter status lights 'Operational test

h) Radiation monitor alarm Opeational test

i) Radiation monitor andlor alarm' Operational test

enabled upon opening of shield door

j) Izitercoms Operational test

In addition to the above, he medical therapy facility minor scrarm shall be tested prior

to reactor startup if the reactor has been shut down for more than sixteen hours.

13. Manual operation of the medical therapy facility's shield door in which the door is

opened fully shall be verified sermi-annually. -

14. Use of thee medical'therapy facilitj beam shall be subject to the following:

a) Atcalibration check of the beam and a functional check of the beam monitors

that are described in provisioufi I Iof this specification shall be iiade weekly for

- any week that the beam will be used for humnan therapy. These checks shall be

*made'prir to any patient irradiation fora given week. In addition, a calibration

*check shall be' perforrned pior to any patient irradiation in the'event that any

cornponent of a given bearn design has been replaied. Finallya calibration

and a functional check shall be perfomed prior to any patient irradiation in the

event of a design modification.

-b) -A characterization of the'bearn shall be performed every si*ponths for any six-

* month interval that the beam'will be usted 'for human' therapy.. This six-month

characterization shall be made prior to'any pitient'irradiation for a given six-

'month interval. A characterization shall also be perfofrned prior to any patient

irradiation in the event of a design modification.- As part of the characterization

process.'the proper response "of the beam monitors that are described in

provision II of this specification shall be verified.

Amendment No. 27

6-25
c) A calibration of the beam monitors that are described in provision II of this

specification shall be performed at least once every two years for any two-year

interval that the beam will be used for human therapy. The two-year calibration

shall be made prior to any patient irradiation during any given

two-year interval.

IS. Maintenance, repair, and modification of the medical therapy facility shall be

performed under the supervision of a senior reactor operator who is licensed by the

U.S. Nuclear Regulatory Commission to operate the MIT Research Reactor. The

'medical therapy facility' includes the beam, beam shutters, beam monitoring

equipment, medical therapy facility shielding, shield door, and patient viewing

equipment. AU modifications will be reviewed pursuant to the requirements of 10

CFR 50.59. The operating couch, patient positioning equipment, medical

instruments, and other equipment used for the direct medical support of the patient

are not considered part of the medical therapy facility for purposes of this provision,

except insofar as radiation safety (i.e., activation and/or contamination) is concerned.

16. Personnel who are not licensed to operate the MIT Research Reactor but who are

responsible for either the medical therapy or the beam's design including construction

and/or modification may operate the controls for the medical therapy facility beam

provided that:

(a) Training has been provided and proficiency satisfactorily demonstrated on the

design of the facility, its controls, and the use of those controls. Proficiency

shall be demonstrated annually.

(b) Instructions are posted at the medical therapy facility's local control panel that

specify the procedure to be followed:

(i) to ensure that only the patient is in the treatment room before turning the

primary beam of radiation on to begin a treatment;

(ii) if the operator is unable to turn the primary beam of radiation off with

controls outside the medical therapy facility, or if any other abnormal

Amendment No. 27

6-26

'\
condition occurs. A directive shall be included with these instructions

to notify the reactor console operator in the event of any abnormality.

(c) In the event that a shutter affects reactivity (e.g., the D2 0 shutter), personnel

who are not licensed on the MIT Research Reactor but who have been trained

under this provision may operate that shutter provided that verbal permission

is requested and received from the reactor console operator immediately prior

to such action. Emergency closures are an exception and may be made

without first requesting permission.

* Records of the training provided under subparagraph (a) above shall be retained in

accordance with the MIT Research Reactor's training program or at least for three

years. A list of personnel so qualified shall be maintained in the reactor control

room.

17. Events defined as 'recordable' under definition 8 of this specification shall be

- recorded and the record maintained for five years. Events defined as

'misadministrations' under definition 9 of this specification shall be reported to the

U.S. Nuclear Regulatory Commission (24 hours verbal. 15 day written report). The

24 hour verbal reports will be made to the Regional Administrator, Region I, or his

designate. The 15 day written reports will be sent to the NRC Document Control

Desk with a copy to the Regional Administrator, Region L or his designate.

18. The requirements of the Quality Management Program (QMPj for the Generation of

Medical Therapy Facility Beam for Human Therapy at the Massachusetts Institute of

Technology Research Reactor shall be observed for any human therapy. (Nlo: The

presence of this commuitment to observe the QMP in these specifications does not

preclude modifying the QMP.as provided in that document. Any such modifications

are not considered to be a change to the MITR Technical Specifications.)

..................................................

Amendment No. 27

6-27
Definhitins

1. The medical therapy facility is equipped with shutters that are used (i) to control beam

delivery and (ii) to adjust the neutron energy spectrum of the beam. The former

currently include lead, boral, and light water shutters as described in Reference

6.5-1. The heavy water blister tank, which is also described in Reference 6.5-1, is

an example of the latter. It is conceivable that these designations may change should

it be found desirable to alter the beam configuration. Accordingly, the phrase

"shutters that control beam delivery" refers either to the aforementioned three existing

shutters or to any future shutter or group thereof that provides an equivalent or

greater reduction in beam intensity. Shutter-effect analyses shall be documented

through the standard safety review process including, where appropriate, an SAR

revision and submission to NRC under 10 CFR 50.59.

2. The term 'calibration check' refers to the process of checking the beam intensity and

quality via one or more of the following: foil activation; use of a fission chamber,

use of an ion chamber, or an equivalent process. The purpose of a calibration check

is to ensure that the beam has not changed in a significant way (e.g., energy

spectrum or intensity) from the beamn that was characterized.

3. The term 'functional check of the beam monitors' shall consist of verifying that

system output is consistent (± 10%) with previously measured values upon

normalization to a common reactor neutronic power level. Ad

4. The term 'charactenzation' refers to the process of obtaining the dose-versus-depth

profile in phantoms as described in Reference 6.5-2 or an equivalent process. The

dose-versus-depth profile from the surface of- the phantom to a depth at least

equivalent to the total thickness of the body part to be treated on a central axis is

deemed adequate for a characterization. Fast neutron, thermal neutron, and gamma

ray components are determined in a characterization and monitors are normalized by

this characterization.

Amendment No. 27

6-28

' >
5. The term 'calibration of the beam monitors' refers to the process whereby the beam

monitors that are described in provision 11 of this specification are calibrated against

instruments that measure dose including a tissue-equivalent chamber and a graphite

or magnesium wall ionization chamber (or the equivalent to any of these three) that

have in turn been calibrated by a secondary calibration laboratory.

6. The term 'design modification' as applied to the medical therapy facility beam refers

(a) to a change that is shown to alter the dose-versus-depth profile of the fast

neutrons, thermal neutrons, or gamma rays in the beam as sensed by the calibration

check and (b) to a change that has the potential to increase significantly the amount of

activation products in the medical therapy facility when the beam is to be used for the

treament of human patients.

7.' Theterm 'radiation fluence' means the total fluence of neutrons and gamma radiation

that is enitted in the medical therapy facility beam. The detewrination of the ratios of

gamma, fast neutron, and thermal neutron fluences is part of the beam

characterization. Knowledge of these ratios allows the total radiation fluence to be

monitored by the on-line detectors, which are neutron-sensitive. Compliance with

the'limits specified on radiation fluence by this specification is determined by

reference to the fluence monitored by these detectors.

8. The term 'recordable event' means the adminison of:

(a) A radiation treatment without a written direciive; or

(b)' A radiation treatment where a written directive is required without reporting to

the medical use licensee in writing each fluence given within 24 hours of the

treatneit or

(c) A treatment delivery for which the administered radiation fluence for any

given'fraction is 15% greater than prescribed.

9. The term 'nisadrninistration' meansthe iadminisration of a radiation therapy:

(a) 'involving the -,vrong patient, wrong mode of treatment, or wrong treatment

.. ~ ..~ , . . -, .: . . .:. .- , . - -

site; or

Amendment No. 27

6-29
(b) When the treatment delivery is not in accordance with provision II of this

specification.

10. The term 'written directive' means an order in writing for a specific patient, dated and

signed by a physician authorized user prior to the administration of radiation and

which specifies the treatment site, the total radiation fluence, radiation fluence per

fraction, and overall treatment period.

11. The term 'human therapy' means radiation treatments that are of direct therapeutic

benefit to the patient and/or part of investigatory studies that involve humans.

12. The term 'physician authorized user' means a medical physician approved for

neutron capture therapy by an NRC-approved medical use licensee.

13. The term 'certified medical physicist' means a medical physicist certified in either

radiological physics or therapeutic radiation physics by the American Board of

Radiology, or in therapeutic radiation physics by the American Board of Medical

Physics and who also has specific training in neutron dosimetry and neutron beam

capture therapy.

The stipulation that patients only be accepted from NRC Medical Use Licensee No.

20-03857-06 or from any other medical use licensee that has been similarly authorized by

NRC to utilize the MIT Research Reactor's Medical Therapy Facility beam for human

therapy, ensures that medical criteria imposed by NRC on such licensees for the use of the

MIT Research Reactor's medical therapy facility beam for human therapy will be fulfilled.

The second provision delineates the division of responsibilities between the Massachusetts

Institute of Technology and the medical licensee that refers the patient. Also, it establishes

administrative authority and protocol for initiating and terminating a radiation therapy.

The requirement that it be possible to initiate a minor scram from a control panel

located in the medical therapy facility area assures the attending physician and/or medical

physicist of the capability to terminate the treatment imnediately should the need arise. The

provision that access to the medical therapy facility be limited to a single door ensures that

Amendment No. 27

6-30

K>~
there will be no inadvertent entries. The various interlocks for the shutters that control beam

delivery ensure that exposure levels in the medical therapy facility will be minimal prior to

entry by personnel who are attending the patient. The shutter-indication lights serve to notify

personnel of the beamds status. The provision for a radiation monitor ensures that personnel

'will have information available on radiation levels in the medical therapy facility prior to

entry. The purpose of this monitor's audible alarm is to alert personnel to the presence of

elevated radiation levels, such as exist when the shutters that control beam delivery are open.

This monitor and/or its alarm may be disabled once the medical therapy facility has been

searched and secured so that it will (1) not disturb a patient and (2) not distract attending

personnel. The monitor and/or its alarm are interlocked with the shield door so that they are

made functional upon opening that door, and hence prior to any possible entry to the medical

therapy facility. One intercom provides a means for the prompt exchange of information

between medical personnel and the reactor operator(s). The second intercom is for

monitoring the patient.

The provision for manual operation of the medical therapy facility's shield door

ensures access to any patient in the event of a loss of electrical power. The presence of the.

viewing window and a closed-circuit TV camera provide the attending physician authorized

user and/or medical physicist with the opportunity to monitor the patient visually as well as

through the use of vanous instruments. The viewing window will function even during an

electric power failure because of the provision for emergency lighting

The specificauon that the total radiation fluence for a therapy (i.e., the radiation

fluences for the sum of all fractions specified in a given treatment plan) not exceed that

prescribed in the patient treatment plan by 20% establishes a trigger limit on the delivered

fluence above which NRC has to be notified of a misadministration. The 20% criterion is

based on the definition of misadrnxnistration (clause 4(iv)) as given in 10 CFR 35.2. The

criterion that the difference between the administered and prescribed fluence for any seven

consecutive days is set at 30%. This is also in accordance with the definition of

misadministration (clause 4(iii)) as given in 10 CFR 35.2. Finally, if a treatment involves

Amendment No. 27

6-31
Atthree or fewer fractions, then a more stringent criterion, 10%, applies to the difference

between the total radiation fluence for a therapy and that prescribed in the areatment plan (10

CFR 35.2(4ii)). The -surveillance requirements for beam calibration checks ,and

characterizations provide a mechanism for ensuring that the medical therapy facility and its

beam will perform as originally designed. Similarly, the surveillance requirements on the

beam monitors ensure that these instruments are calibrated by a means traceable to the

National Institute of Standards and Technology. The chambers specified (tissue-equivalent,

and graphite or magnesium-wall) were chosen because they measure dose as opposed to

fluence.

The specification on maintenance and repair of the medical therapy facility ensures

that all such activities are performed under the supervision of personnel cognizant of quality

assurance and other requirements such as radiation safety. The provision on the training and

proficiency of non-licensed personnel ensures that all such personnel will receive instruction

equivalent to that given to licensed reactor operators as regards use of the medical therapy

facility beam. (Note: Licensed reactor operators may, of course, operate the medical therapy

facility beam.) Also, this provision provides for the posting of instructions to be followed in

the event of an abnormality.

The specification on 'recordable events' and 'misadministrations' provides for the

documentation and reporting to the U.S. Nuclear Regulatory Commission of improper

events regarding the generation and use of the medical therapy facility beam. The

requirement that the Quality Management Program (QMP) be observed ensures that radiation

treatments provided by the medical therapy facility beam will be administered as directed by

the physician authonzed user.

Amendment No. 27

6-32

K>
Rcfcrxnc

6.5-1 MITR Staff, "Safety Analysis Report for the MIT Research Reactor (MITR-II),"

Report No. MITNE-115, 22 OCL 1970, Section 10.1.3.

6.5-2 Choi, RJ., "Development and Characterization of an Epithermal Beam for Boron

Neutron Capture Therapy at the MITRA- Research Reactor," Ph.D. Thesis, Nuclear

Engineering Department, Massachusetts Institute of Technology, April 1991.

Amendment No. 27

6-33
(i) Gaseous Waste (Summarized on a monthly basis)

(a) Radioactivity discharged during the reporting period (in

curies) for.

(I) Gases

(2) Particulates, with half lives greater than eight days.

(b) The MPC used and the estimated activity (in curies)

discharged during the reporting period, by nuclide, based on

representative isotopic analysis.

(iii) Soi at

(a) The total amount of solid waste packaged (in cubic feet).

(b) The total activity and type of activity involved (in curies)..

(c) The dates of shipment and disposition (if shipped offsite).

i. A summary of the use of the medical therapy facility for human therapy.

(i) Investgative Studies

(a) Nature and status of the studies.

(b) Number of patients involved.

(ii) Human nu

(a) Number of patients Beated

(b) Type of cancer treated.

Amendment No. 27

7-32
UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON. D r. 20555

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

,, . _, A.M. . T - . -N',!

.* . SUPPORTING AMENDMENT NO. 27 TO . ',

. t . .

.1 I I

FACILITY OPERATING LICENSE NO. R-37

DOCKET NO, 50-20

MASSACHUSETTS INSTITUTE OF TECHNOLOGY .

1.0 .INTRODUCTION . . . . . ..

By letter dated Harch 10, 1992, as supplemented on July--30, .1992, August 31,

.1992, September 23, 1992, December 22, 1992, December 30,1992, and

January 22, 1993, the'Massachusetts Institute of Technology (MIT) requested

changes to the Technical Specifications.,(TS) ofFacility Operating License

No. R-37 for the HIT Research Reactor.(MITR-I1).. The-requested ,changes would

add requirements to the TS concerning the use of the MITR-11 Medical Therapy

beam (beam) for human therapy.

2.0 BACKGROUND .

The Atomic Energy Act of 1954,-as amended-(the Act) authorizes the issuance of

--class 104 -licenses for redical.therapy (Section 104(a) of the Act) andjfor,

: research and development facilities (Section 104(c) of the Act)... All non-i

power reactors (NPRs) currently licensed by-the U. S. Nuclear Regulitory..-

-:Commission (NRC or-the Commission) hold class 104c licenses in accordance.with

10 CFR 50.21(c). HNIT also holds a 104a license in accordance with 10 CFR..

:50.21(a). HIT is the only NPR licensee to hold a 104a license. -This license

was issued in 1958 when-the original.MIT reactor was lieqnsed by the Atomic

Energy Commission. : - -

-A potential medical therapy use for NPRs thatwas identifiedin.the 1950s-is

the use of 'neutron beams to treat,the cancers glioblastoma multiform and

metastasized melanoma. Called Boron Neutron Capture Therapy (BNCT), the

treatment consists of a patient ingesting a boron compound that concentrates

boron in the tumors. The patient is then exposed to a neutron beam'from the

NPR which causes the boron to fission into lithium.and an alpha particle,.

which are heavy charged particles.;, These.particles cause.secondary ionization

that kills the tumor cells.

MIT conducted patient trialsduring 1960-1962.with no success.--.Apparently,

the boron drugs available at that time did not-concentrate enough -boron in the

tumor and the use of thermal neutron-beams did not:allow neutrons to reach

deep enough into the brain for successful treatment. '
-2-

Since those early trials, improved boron compounds have been developed that

are thought to increase concentrations of boron in the cancer cells. Also,

MIT has focused on the use of an epithermal neutron beam that will allow

neutrons to reach any part of the brain. These two developments have led.

researchers to believe that another attempt should be made to treat cancers

using BNCT.

In letters to the NRC staff dated April 3, April 16, and June 6, 1991, MIT

outlined the current proposed practice and procedures for using neutron beams

in patient therapy, the history of the 1960-1962 patient trials, and

historical information on the procedures for the medical beam therapy facility

at MIT. MIT also expressed the opinion that they satisfy all NRC regulations

concerning the use of neutron beams'for the treatment of human subjects. The

existing regulations authorize the issuance of licenses for human therapy but

places no requirements or restrictions on the irradiation of humans with

neutron beams from NPRs.

During the thirty years since the MIT patient trials, NRC has issued no other

104a license for NPR medical therapy. Thus, the NRC has not developed

regulations for the use of special nuclear material for medical therapy at

NPRs. However, regulations for the medical use of byproduct material have

been developed in 10 CFR Pirt 35, Medical Use of Byproduct Material."

Although 10 CFR Part 35 may be used as a guide to regulate BNCT, it is-not

directly applicable to BNCT and the use of special nuclear material for human

therapy.

The use of neutron beams for medical therapy has many similarities to

teletherapy. Because of this, the NRC staff decided to use the criteria of

10 CFR Part 35, Subpart 1, Teletherapy, as a model for the regulation of

neutron beams for the treatment of human patients. -The'requirements of Part

35 were modified to account for the differences between teletherapy and BNCT.

For regulatory purposes, BNCT is divided into two components: beam generation

and beam use. MIT is responsible for the regulatory aspects of beam

generation and the hospital providing treatment to the patient is responsible

for the regulatory aspects of beam use. A February 19, 1992, letter to MIT

from NRC outlined the commitments and information that MIT was requested to

submit to NRC. Because no 10 CFR Part 50 regulations that parallel 10 CFR

Part 35 exist, the commitments made by MIT for the conduct of human therapy

would be captured in the facility Technical Specifications. The license

amendment request from MIT evaluated by the staff in this safety evaluation

report (SER) is in response to the February 19, 1992, NRC letter.

3.0 EVALUATION

The commitments and information requested by NRC of MIT to be included in the

request for license amendment are as follows:

(I) A commitment to limit'the delivery of neutrons only to human subjects

pursuant to a',written directive from a physician authorized user who is

specifically authorized to perform boron neutron capture therapy by NRC

medical use licensee No.'20-03857-06 (New England Medical Center).
- 3 -

(2) A commitment to.record events equivalent to recordable events' in

10 CFR 35.2, report events equivalent to "misadministrations" in

10 CFR 35.2,- and establish a written quality management program using:

the criteria specified in 10 CFR Part 35 for teletherapy (the neutron-

-,,beam).

(3) The methodology to ensure that the neutron flux, fluence, and spectrum ' delivered to the patient are as requested by the physician authorized

user.

(4) The design aspects of the neutron beam delivery system that are-'

important to patient or user safetysuch that these aspects cannot be

changed without license amendment.

(5) The reactor operator and physician communication system and the method

for terminating the treatment exposures.

(6) A list of the anticipated activities that may alter beam characteristics

and may require spot-checks before the beam use and th'e spot-checks that

will be performed in these situations.

(7) ,- The interlock systems and safety precautions used to prevent personnel ', - from being accidentally exposed to the beam in the treatment-room and'.

the safety precautions to be followed before, during, and after.'_

treatment exposures to limit occupational exposure to ionizing

radiation. Include information-on sur'veillances,,if required.

Each of these areas have been addre'ssed' bythe licensee in their' amendment

request.

A new section-to the TS,.Section 6.5, has been proposed by the licensee to,

place limits''ind conditions on the generation of a neutron beam for human ' therapy. These TSLwill ensure that patients are-treated in accordance with'

the treatment plan established by the physician authorized user,'tthat the as

low as'reasonably achievable (ALARA)'principle.is observed.for all nontherapeutic'r.adiation'exposures, and that the'medical'therapy treatment room

design features'operate properly. '

The licensee is not seeking approval of-the details of beam design.. Except'.

for the initial facility design, and the review of changes to the-facility,.'

experiments, and'procedures condicted.under' 10'CFR'50.59'for.the beam and

submitted toNRC, the'NRC'stiff has not.reviewed the'specific'design of the, . beam filter.systemor the gamma and,'neutron fluxes and energies of.the beam.".'

The ability to place'filters 'in the beam to control beamncharacteristics is''

allowed in the existing'license and,'is''not unique to medical therapy. The

filter construction and materials must beWin accordance with 'the.license and

safety analysis report or changes'approved.under.10,CFR 50.'59. "The'characteristics of.the beam necessary'to conduct -treatment-are decisions that are

medical in nature and"are the"responsibility of'the physician authorized'user.

The NRC staff is also reviewing an amendment to the medical use of byproduct

material license for New England Medical Center to regulate beam use.
- 4 -

TS 6.5.1 requires that patients be referred by a written directive of the

physician authorized user from a medical use licensee authorized by NRC to

utilize the MITR-I.' TS 6.5.2 clearly defines that treatment is the

responsibility of the physician authorized user in charge of the therapy and

that delivery of the requested radiation fluence is the responsibility of MIT.

These TS are responsive to item (1) above.

TS 6.5.2 also assigns responsibility to medical and MIT personnel for the

initiation, monitoring, and termination of treatment. TS 6.5.10 requires that

the patient be observed through both a viewing port in the medical therapy

facility and by a closed-circuit TV camera.' Therapy may continue at the

discretion of the physician authorized user if one of the viewing methods

fails during the treatment. These TS are responsive to item (5) above. The

staff concludes that patients will be properly referred for treatment and that

the responsibility for the treatment will be properly controlled. This is

acceptable to the staff.

TS 6.5.3 requires that' it be possible to initiate a minor scram of the reactor

from the medical therapy room control panel and provides a temporary provision

of communicating with the control room operator to'scram the reactor if the

minor scram button on the medical therapy room control panel is out of

service. This temporary measure is limited to seven consecutive working days.

TS 6.5.8 requires that an intercom or other mean's of two-way communication

exist between the medical' therapy facility control panel and the control room

and between the medical 'therapy facility control panel and the inside of the

therapy room. This will'allow the reactor, which is the source of neutrons,

to be rapidly shut down in an emergency. These TS are responsive to items

(5) and (7) above. The staff concludes that therapy can be rapidly terminated

if the need arises. This is acceptable to the staff.

Item (7) above focuses on interlocks and safety precautions to prevent Ks"

accidental personnel exposure and surveillances on these interlocks. TS

6.5.4, 6.5.5, 6.5.6, and 6.5.9 describe the operation and'interlocks on the

medical' therapy facility doors that'allow access into the facility and on the

shutters that control beam delivery and, when closed, shield the inside of the

medical therapy facility room from the reactor. TS 6.5.4 requires that access

to the facility be controlled by a shield door. TS 6.5.5 requires that an

interlock prevent the shutters from opening unless the facilit? shield door is

closed. If the shutters are open and the shield door is opened, the shutters

are interlocked to automatically close. The shutters will also close

automatically if electric power fails. If the shutters are pneumatically

powered and air pressure becomes low, the shutters will automatically close.

The shutters are also-designed to be manually-closed. TS 6.5.5 also requires

that the shutters are able to be closed manually from within the medical

therapy facility. TS 6.5.6 requires that the status of the'shutters be

indicated by lights. -A temporary alternate means of indicating' shutter

position may be used for up to seven consecutive working days if the light

system fails. TS 6.5.9 'requires that the facility shield door be able to be

opened manually.
TS 6.5.12 and 6.5.13 requires surveillance tests of the above interlocks and

the medical therapy facility-minor reactor scram.- When the facility is in'use - for human therapy, -the surveillances are conducted monthly with the exception

of the full opening manual operation of the shield door which is semi-annual.

These surveillances will also be conducted'ifrepairs have occurred or'the''-

system has been deenergized. In addition,-if the'reactor has been shutdown

for more-than sixteen hours, -the medical therapy facility minor scram shall be

tested.:

The staff has determined that these TS requirements ensure that adequate

safety precautions exist to prevent accidental exposure from the medical

therapy facility beam and that surveillances exist to test the interlocks'and

scram. This is acceptable to the-staff.,

TS 6.5.7'concerns the requirements for:radiation safety and monitoring of the'

medical therapy facility. -This TS is also'responsive to''item (7) above. The'

TS requires a radiationmonitor in the medical therapy facility'with visual..

readout and local alarm. -The primary purpose 'of this monitor is to warn the

MIT staff if the shutters are not properly closed. The TS also encompasses

requirements for backup-power, checks for proper operation and -quarterly

calibration. ..The TS-places the monitor setpoint at or below 50 mr/hr. This

radiation setpoint is above the radiation levels in the room immediately after

termination of treatment, but is low enough to' provide warning if the-shutters

are not properly closed. The licensee'has the'ability,-in accordance with the

TS, to disable -the monitor during therapy. -This will prevent'alarming of the

monitor'during therapy sessions which would-be a-distraction -to the patient

and the treatment staff. If the monitor is disabled,-it'shall be interlocked'

to automatically become functional again upon-opening'of the shield door.'.

This TS-also, allows for temporary means-of meeting the monitor requirements -

for up to seven-consecutive-working days with portable-survey meters'or

audible dosimeters if the monitor is inoperable. Surveillaince'requirements

for quarterly calibration of the radiation monitor are included in the TS. '

The staff concludes that the medical therapy facility will be adequately

monitored for radiation levels. -This is acceptable-to the staff. '' '

TS 6.5.11 concerns the accuracy of the radiation fluence deliVered by the '' -

medical therapy facility.beam. - This is responsive to item:.(2) above. The : ' ' total fluence delivered-shall not-exceed the prescribed amount by more than ~ - '

20 percent. There are also criteria for the administered and prescribed -'

fluence not to exceed 30 percent for any given week (seven consecutive days)

and a 10 percent-criteria if the treatment consists of three or fewer' -

fractions. Recordable events and misadministrations are defined -in the' ' - -

definitions-section-ofTS 6.5.-.,TS-6.5.17 -has the requirements 'for maintaining' A'

records of recordable eventsiand for reporting misadministrations to the'NRC.

Recordable events records shallbbe maintained for'five years. Kisadministrai' '--' tions shall-be verbally reported-to;NRC within 24-hours and in writing within

15 days. The definitions-of recordable events-and misadministrations agree'

with the definitions given in:10 CFR Part 35 for teletherapy. This section of

the TS is acceptable.totthe-staff.-. l; - -
- 6 -

TS 6.5.14 has the requirements-for calibration checks, characterization, and

calibration of the beam monitors for the medical therapy'facility beam. The

beam' is calibration checked and the beam monitors are functional checked

weekly when the beam is in use for treatment prior to treatment for that week.

The beam'will be characterized every-six months for any six month interval

that the beam will be used for human therapy prior, to patient irradiation for

the six month interval. If beam design modification occurs, the beam shall be

calibration checked, functional checked, and characterized before patients are

irradiated. In addition, a calibration check will be performed in the event

of a beam component replacement.

The beam shall be calibrated against dose measuring instruments that have been

calibrated by a secondary laboratory at least once every two years for any two

year interval that the beam will be used for human therapy prior to patient

irradiation for the two year interval. Calibration-checks' characterization,

functional checks, and calibration are defined in the definition section of

TS 6.5. These provisions of the TS are responsive to items (3) and (6) above

and will ensure that the beam characteristics remain stable over time.

The staff has determined that the calibrations, beam characterizations,

functional checks, and calibration checks required by the TS are acceptable.

TS 5.6.15 controls maintenance, repair, and modification of the medical

therapy facility. These changes to the facility will be made under the

direction of an NRC licensed senior reactor operator. The TS confirms that

changes to the facility will be reviewed pursuant to 10 CFR 50.59.' The TS

also states that certain items such as' the operating couch, patient

positioning equipment, medical instruments, and other equipment used for the

direct medical support of the patient are not considered part of the medical

therapy facility for the purposes of maintenance, rEpair or modification.

These are the responsibility of the medical provider. However, MIT is

responsible for the radiation safety aspects (activation and/or contamination

control) of this equipment.

This TS responds to item (4) above. The staff concludes that maintenance,

repair, and modification of the medical therapy facility will be adequately

controlled, the NRC will be involved at the appropriate times through

compliance with 10 CFR 50.59, and that MIT will be responsible for radiation

safety of equipment in the medical therapy facility. The staff finds this TS

to be acceptable.

TS 5.6.16 has the requirements for the training of personnel who are not

reactor operators but who are responsible for medical therapy, beam design, or

construction and/or modification of the beam. Instructions shall be provided

to ensure that only the patient is in the therapy facility prior to the

initialization of treatment, and for the steps to be taken if abnormal

conditions occur or if the medical therapy facility controlsfail to turn off

the beam. Training records will be maintained for three years or in

accordance with the reactor training program, and a list of qualified

personnel will be maintained in the reactor control room. There is a
- 7 -

requirement that, except in emergences, manipulating shutters that could

affect the reactivity of the reactor must be authorized in advance by the

reactor console operator.

"The staff has determined that personnel that operate the medical therapy - facility will be properly trained in the use of the facility and that except

in emergencies, the reactivity of the reactor will only be manipulated with

the permission of the control room operator.

TS 5:6.18 'requires that the Quality Management Program (QMP) for the - Generation of Medical Therapy Facility Beam for Human Therapy at-the i

Massachusetts Institute of Technology Research Reactor'(Attachment l) be

observed for'human-therapy'. The program parallels the requirements of 10 CFR

35.32, Quality Management Program. The program meets objectives similar to.

those of 10 CFR 35.32 for having a written directive prior to administration

of treatment, for verification of the patient's identity by more than one

'method prior to treatment, that the treatment plan and related calculations

are inaccordance with the written directive,, that the treatment is in

.accordance with the written directive,'that unintended deviations from the

written directive are identified and evaluated, and that appropriate action is

taken.-

Although the QMP is a requirement of the TS, the QMP is not part of the TS and

may be changed by the licensee without license amendment. The requirementsfor QMP modifications are given in the QMP. - -

The licensee will develop written procedures for the conduct of therapy andwill provide them to the NRC at least thirty days prior to the start of human

irradiations.

The staff has reviewed the QMP and finds that there is assurance that the

objectives discussed above will be met by MIT. This section of the TS is

acceptable to the staff.

TS 7.13.5 on reporting requirements is amended to add a requirement that a

summary of the use of the medical therapy facility for humaq therapy be

included in the annual report of the MITR-I1. The staff finds this addition'

to the annual report to be acceptable.

3.0 ENVIRONMENTAL CONSIDERATION

This amendment involves changes in the installation or use of facility

components located within the restricted area as defined in 10 CFR Part 20 and

changes in inspection and surveillance requirements. The staff has determined

that the amendment involves no significant increase in the amounts, and no

significant change in the types, of any effluents that may be released

offsite, and there is no significant increase in individual or cumulative
ALoccupational radiation exposure. Accordingly, this amendment meets the

eligibility criteria for categorical exclusion set forth in 10 CFR

51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement

or environmental assessment need be prepared in connection with the issuance

of this amendment.

4.0 CONCLUSIOT

The staff has concluded that there is assurance that the use of the MIT

medical therapy facility-and beam for human therapy-will be in accordance with

the treatment plan established by the physician authorized user, that nontherapeutic radiation exposures will be ALARA, and that the medical therapy

facility will function as designed.

The staff has also concluded, based on the considerations. discussed above,

that: (1) because the amendment does not involve a significant: increase in the

probability or consequences of reactor accidents previously evaluated, or

create the possibility of a new or different kind of reactor accident from any

reactor accident previously evaluated,.and does not involve a significant

reduction in a margin of safety, the amendment does not involve a significant

reactor hazards consideration, (2) there is reasonable assurance that the

health and safety of the public will not be endangered by. the proposed

activities, and (3) such activities will be conducted in compliance with the

Commission's regulations and the issuance of this amendment will not be

inimical to the common defense and security or the health and safety of the

public.

Principal Contributors: Alexander Adams, Jr.

Donna-Beth Howe

James A. Smith

Attachment:

As stated

Date: February 16, 1993
Ouality Management Program

for

Generation of MITR-fl Medical Therapy Facility Beam

for Human Therapy

ATTAtT 1
Ou~iryMangerrent roijm: Generatio'nof MTTR-TT Medical Ther2arvPciliry Beam

for Human Th-raov

1. - Pupose: The objective of this quality management program is to ensure that

radiation treatmnents provided by the MrT Research Reactor's (MITR-HI) Medical

Therapy Facility beam will be administered as directed by a physician authorized

.user.

2. Authorized Medica Use Licensees: Use of the MIT Research Reactor's Medical

Therapy Facility beam, for the treatment of human subjects, is limited to the

physician authorized users authorized under

(a) NRC Medical Use Licensee No. 20-03857-06.

(b) Any other medical use licensee that has been similarly authorized by NRC to

utilize the MIT Research Reactor's Medical Therapy Facility beam for - . human therapy.

3. Program Requirements: The following requirements are established as part of this

quality management program:

(a) A written directive will, except as noted in subparagraph (iv) below, be

-prepared by a physician authorized user of the NRC-approved medical use

licensee prior to the administration of any radiation therapy. This directive

shall be written, signed, and dated by the physician authorized user and it

shall include the following information:

(i) Name and other means of identifying the patient

(ii) Name of the physician authorized user and certified medical -. physicist in charge of the therapy.

(iii) The total radiation fluence to be administered, the radiation fluence

per fraction, the treatment site, and the overall treatmaet period.

(iv) If, because of the patient's condition, a delay in order to provide a

written revision to an existing written directive would jeopardiii the

patient's health, an oral revision to an existing written directive will

be acceptable, provided the oral revision is'dociented immediately - in the patient's record and a revised written directive is signed by a - physician authorized userwithin 48 hours of the oral revision.

Also, a written revision to an existing written directive may be made

-for any therapeutic procedure provided that the revision is dated and

signed by i physician authorized user prior to the administration of

the next fraction.

If,--because of the emergency nature of the patient's condition, a

delay in'order to provide a written directive would jeopardize the

patient's health. an oral directive will be acceptable; provided that

the information contained in the oral directive 'is documented

immediately in the patient's record and a written directive is prepared

within 24 hours of the oral directive. - - - -

(v) In order to ensure that the Staff of the MIT Research Reactor has the

most recent written direcuve from the medical use licensee and the
-2-

correct directive for the patient in question, a copy of that directive

shall be handdelivered to the MMTR Staff by the Staff of the medical

use licensee who accompany the patient to MIT. This copy shall

then be checked against the most recent previous transmission. Any

discrepancy shall be resolved by the medical use licensee puior to the

initiation of patient irradiation.

(vi) The Director of the MIT Nuclear Reactor Laboratory, or his

designate, will date and sign the written directive to verify that

current and accurate beam characteristic parameters were provided to

the NRC-approved medical use licensee and that the radiation

fluence desired in the written directive was delivered. A copy of this

signed directive shall be provided to the medical use licensee within

twenty-four hours of a treauenc.

(b) Prior to each administration of any radiation, the patient's identity will be

verified by more than one method as the individual named in the written

directive. The Mrr Nuclear Reactor Laboraorcy will use any two or more of

the following acceptable methods of identification:

6X) Self-identification by patients who are conscious upon arrival at the

MIT Research Reactor. Information provided by the patient shall

include any two of the following. name, address, date of birth, or

social security number. The information provided by the patient is

to be compared to the corresponding information in the patient's

record.

(ii) Hospital wrist band identification with the wrist band information to

be compared to the corresponding information in the patient's

record.

(iii) Visual identification against photographs provided with the written V

directive.

(iv) Other methods as specified in U.S. Nuclear Regulatory Commission

Regulatory Guide 8.33, "Quality Managemnent Program."

(c) The plan of treamient is certified by the certified medical physicist to be in

accordance with the written directive. In this regard: the Massachusetts

Institute of Technology is responsible for calibrating the output of the beam

monitoring instrumentation versus dose in phantom and for providing a

central axis dose versus depth profile. This information will then be used

by personnel at the NRC-approved medical use licensee to generate a plan of treatmrent - Conformance of the beam to its design characteristics is

confirmed through the measurements specified in MITR Technical

Specification #6.5, "Generation of Medical Therapy Facility Beam for

'Human Therapy." The beam is characterized dosimetrically every six

months (provision 14(b)), the beam monitors are calibrated every two years by a secondary calibration laboratory and their proper operation is verified

semi-annually (provision 14(c)), and calibration checks are made of the

beam at least weekly for any week that the beam will be used for human

therapy (provision 14(a)).

(d) Each administration of radiation is in accordance with the written directive

subject to the tolerances established in provision 11 of MMR Technical

Specification #6.5, "Generation of Medical Therapy Beam for Human

Therapy."
-3 -

(e) Any unintended deviations from the written directive shall be identified and

evaluated,'and appropriate action taken. Such action shall include informing

the medical use licensee'of the 'deviation.' These reviews shall be performed

monthly for any month in which human therapy was conducted. For each

patient case reviewed, it shall be determined whether the administered total - 'fluenc fluence per fraction, treatment site,' arid overall treatmeni period

were as Specified in the written directive. In the event of any deviation from

the written directive, the licensee (MIT) shall identify its cause and the

action required to prevent recurence. These actions may include new or

revised policies, new or revised procedures, additional training, increased

supervisory review of work, or other measures as deemed appropriate. Corrective actions shall be implemented as soon as practicable.

4. '`Pro'ranm Implementation: The following practices shall be observed in order to

ensure proper implementation of the quality management program:

(i) tA revieiw shall be conducted of the q'ality management program.' This

review shall include, since the last review, an evaluation of:

(i) A representative sample of patient administrations,

(ii) All recordable events, and

(iii) All nisadministrations.

The objective of this review is to verify compliantce with all aspects of the

quality management program. For purposes of this review, the term

'representative' in statement (i) above is defined as 100% sampling up to

twenty patients; a sample of twenty for twenty-one to one hundred patients,

and 20% sampling for more than one hundred patients. In order to

eliminate any bias in the sample, the patient cases to be reviewed should be

-selected randomly.'

(b) The procedure for conducting the above review is as follows:

-I) -The review shall be performed by the Director of the MIT Radiation

Protecton Program or his designate.

(i) The review shall be performed annually.

(ii) Patient administrations selected for review shall be audited to

determine compliance with each of the requirements listed in

'paragraph (3) above.

(iv) The review shall be written and any items that require further action

shall be so designated. Copies of the review shall be provided to the

NRL Director and to the MIT Reactor Safeguards Committee who

will evaluate each review and. if required, recommend modifications

-in this quality management program to meet the requirements of

paragraph (3) above. A copy of these reviews will also be provided

to each medical use licensee.

(c) - Records 'of each review, including the evaluations and findings of the

review, shall be retained in an auditable fonn for thre years.--' .; . . .- . . : - i .- -
-4-

(d) The licensee (MMI) shall reevaluate the. Quality Management Program's policies and procedures after each annual review to determine whether the

program is still effective or to identify actions required to make the program

more effective.

5. Response to Recordable Event: Within thirty days after the discovery of a

recordable event, the event shall be evaluated and a response made that includes:

(a) Assembling the relevant facts, including the cause;

(b) Identifying what. if any, corrective action is required to prevent recurrence;

and

(c) Retaining a record. in an auditable form, for three years, of the relevant

facts and what corrective action, if any, was taken.

A copy of any recordable event shall be provided to the affected medical use

licensee.

6. Records Retention: The following records shall be retained.

(a) Each written directive for three years; and

(b) A record of each administered radiation therapy where a written directive is

required in paragraph (3(a)) above, in an auditable form, for three years after the date of administration.

7. Program Modificanion: Modifications may be made to this quality management

program to increase the, program's efficiency provided that the program's effectiveness is not decreaied All medical use licensees shall be notified of any modifications and provided with a copy of the revised program. The licensee

(MMT) shall furnish the modification to the NRC (Region I) within 30 days after the

modification has been made.

8. Re=on and Surveillance Feuencv. Any report or other function that is required

to be performed in this Quality Management Program at a specified frequency shall

be performed within the specified time interval with:'

(a) a maximum allowable extension not to exceed 25%'of the specified surveillance interval, unless otherwise stated in this Quality Management

Program;

(b) a total maximum combined interval time for any: three consecutive

surveillance intervals not to exceed 3.25 times the specified surveillance

interval.

9. Definitions

(a) The term 'physician authorized user' means a medical physician approved for neutron capture therapy by an NRC-apprved medical use licensee.

(b) The term 'certified medical physicist' means a medical physicist certified in

either radiological physics or therapeutic radiation physics by the American

Board of Radiology, or in therapeutic radiation physics by the American

Board of Medical Physics and who also has specific training in neutron

dosimetry and neutron beam capture therapy.
.5-

10. sAgflicabilijy: This Quality Management Program applies solely to the generation of the medical therapy facility beam for the ueaunent of human subjects. It does not apply to any other use of the medical therapy facility and/or its beam. Reports nd surveillances listed in this specification are only required if human therapy was conducted during the referenced interval.
17 DECOMMISSIONING AND POSSESSION-ONLY

LICENSE AMENDMENTS

Since the first nuclear reactor was assembled in 1942,' many non-power reactors,

critical facilities, and special-purpose reactors have been dismantled,

decontaminated, and'decomniissioned.' More than 65 non-power reactors'

regiiWed by NRC have been dismantled and their licenses'terminated.'

Approximately the same nuimber of non-power reactors owned by the Department

of Energy'(DOE) or'the Department of Defense have been decomnimissioned.: In

'the absence of detailed regulations governing'decommissioning,'NRC has

developed a systematic approach for licensee and NRC actions to terminate facility

licenses. The approach includes issuance of possession-only license amendments

and orders authorizing facility dismantlement.

On June 27,.1988, NRC published a notice of rulemaking (53 FR 24018) amending

-its existing regulations for terminating licenses. These amendments'affected thedecommissioning of licensed reactor facilities. The amended 10 CER 50.82

requires that an application for a termination of a license be accompanied, or

preceded, by a proposed decommissioning plan (DP). Licensees should be aware

that additional changes have been proposed to the decommissioning regulations'

(see 60 FR 37374). Therefore, the following guidance is offered in the interim to

fiicilitte the review of decommissioning activities during this period.'

17.1 Decommissioning -

17.1.1 Preliminary Decommissioning Plan

In addition to the DP required by 10 CFR 50.82, 10 CFR 50.75(f) requires'each

licensee to submit a preliminary DP. The preliminary DP shall be submitted at or

about 5 years before the projected end of operation. The plan shall contain an

estimate of the cost of decommissioning and an up-to-date assessment of the'major

technical factors that could affect planning for decommissioning. The factors to be

considered in submitting this information include the following:

(1) the decommissioning alternative anticipated to be used, including,

consideration of the requirements of 10 CFR 50.82(b)(1)

(2) major technical activities necessary to carry out decommissioning safely

(3) the current situation for disposal of high- and low-level radioactive waste

(4) residual radioactivity criteria

~R~v0,2/6 1.1 S~i~ FORAT . REV 0, 2/96 - -17-1- -- STAMARD FoRmAT Am Co\-m-\,r
CHAPTER 17

(5) other site-specific factors that could affect decommissioning planning and

cost

If necessary, the cost estimate should also include plans for adjusting levels of

funds assured for dec6mmissioning to demonstrate a reasonable level of assurance

that funds will be available when needed to cover the cost of decommissioning.

The licensee maiy, use the format shown in NRC draft Regulatory Guide DG-1005

in preparing the preliminary DP. The preliminary DP need only address the five

factors listed above and may be substantially less detailed than the final DP.

Additional information on the financial aspects of decommissioning can be found in

Chapter 15, Financial Qualifications," of this format'and content guide.

17.1.2 Decommissioning Plan

Applications for authorization to begin decommissioning and for NRC to terminate

the license shall be filed as required by 10 CFR 50.82. The NRC staff recommends

that the fuel be removed from the core or from an operational configuration under

the existing facility operating license as soon as possible after reactor operations

permanently cease. The fuel should be shipped off site in accordance with DOE,

NRC, and Department of Transportation (DOT) regulations.

The schedule for decommissioning is discussed in 10 CFR 50.82(bXl)(ii), which

requires that a non-power reactor be decommissioned without significant delay,

except when concern for public health and safety makes delay necessary. The

factors to be considered in evaluating an alternative that delays the completion of

decommissioning are discussed in 10 CFR 50.82(bXl(iii). Among the factors that

may warrant a delay in decommissioning are unavailability of waste disposal

facilities; other nuclear facilities, such as another non-power reactor, on the same

site; and other site-specific factors.

Under some circumstances, the licensee can apply for a possession-only license

amendment according to 10 CFR 50.90 after operations have ended and before

decommissioning starts. This amendment normally does not extend the term of the

facility license. Generally, the amendment should be based on the stated intent of

the licensee to develop and submit a DP and an application to dismantle and

decommission the facility and terminate the license. The possession-only license

amendment grants the licensee authority to possess but not operate the facility. If

requested by the licensee and justified by a safety analysis, regulatory relief from

the license and technical specification requirements for an operating non-power

reactor may be authorized with a possession-only license amendment. Further, this

amendment permits the licensee to retain the reactor facility, related radioactive

byproduct material, and, in some cases, special nuclear material, pending approval

of the DP. If a non-power reactor is subject to annual licensing fees, the granting

of a possession-only license amendment also removes the basis for annual fees. v

NUEG157A_ _7. 0.2:~v 

NUREG-1537. PART I 17-2 REV. O. 2/96
DECOMMISSIONING AND POSSESSION-ONLY AmNDMENTs

In 10 CFR 51.53(b), NRC requires that each test reacior licensee submit, in

addition to the DP, a supplement to its environmental report with the application

to decommission. This ER supplement should reflect any information about

significant environmental impacts associated with the proposed decommissioning

activities.

A DP should show that the facility can be dismantled and decontaminated safely by

describing the plans to decontaminate the facility and site to comply with the

release criteria. The DP should describe an organized method for removing

radioactive components and material, reducing contamination and radiation levels

in the facility, 'nd permitting release of the facility; the organization that will

decommission the facility and the process this orgniztion will follow; and the

radiological status of the facility and the method of performing tasks to prevent

undue radiation exposures to the facility staff and the public and the release ofradioactivity to the environment. NRC approves a DP by issuing of an order

authorizing implementation.

Some information necessary to develop the DP may not be available when the DP

is required to be written or changes in the DP may be necessary after

decommissioning begins. Therefore, the DP shoiild provide for accommodation" of

any necessary changes in the'DP and procedures through a process similar to the

one set forth in 10 CFR 50.59. Such a process could ensure that changes to the

plan that are reviewed and determined to be not 'safety significant could be made

with approval of the decommissioning safety committee (or other responsible

licensee official or organization)'and then reported to NRC. 'Safety-significant

changes would still require pnror review and approval by the NRC staff. Because

the DP i"'put into effect with an order, in the absence of a process such as that

described above, the NRC staff would have to review even"muinor changes that

have no safety significance. This could result in the use of licensee and NRC

resources with no concomitant safety benefit.

The NRC staff will not terminate the license until the decommissioned facility

meets release criteria. After the facility has been decontaminated and the

radioactive materials from decommissioning have been disposed of; the licensee

shall measure the residual radiationiand docuiment the measurements in a final

survey report." The licensee shill submit this report to NRC to be evaluated and to

support the request that the license be teirminated. NRC regional stiff or an NRC

contractor will conduct an onsite survey to verify the radiation exposure and

contamination levels documentid in the final survey report. When the

requirements for the release criteria have been satisfied, NRC will terminate the

license.

In some cases, the informiation requested in this format and content guide may be

the same as or similar to infornation previously submitted to NRC. Information

R~v.O,2~9 .73SADR ORA N or

REV. O. 2/96 . .17-3 , - ' 'STANDARD FORMAT AND CON=  .
CHAP= ~17

contained in previous submittals, statements, or reports may be incorporated by

clear and specific references, and only changes need be submitted.

The administrative format of the application and the procedure for submitting the

application should conform to the guidance in Chapter 1 of this format and content

guide.

17.1.3 Decommissioning Alternatives

According to 10 CFR 50.82, two general approaches to decommissioning are

acceptable. Which approach is chosen depends on the type of residual

radioactivity present, the design and number of facilities at a site, the availability of

a waste disposal site, and other factors. One approach is to start decommissioning

activities soon after operations cease and proceed immediately toward terminating

the license and releasing the facility. This is the approach the regulations generally

require for non-power reactors. The other approach is to perform limited

decommissioning activities soon after operations cease to prepare the facility for

safe storage or entombment. The facility will be maintained in a safe condition for

a time, after which the licensee will either decontaminate the facility or allow the

radioactivity to decay to radiation levels that conform with release criteria.

Three decommissioning alternatives are defined in NUREG-0586.

In decontamination (DECON), the equipment, structures, and portions of a

ficility and site containing iadioactive materials are removed or

decontaminated to a level permitting release of the property by NRC

shortly afer operations cease.

* In safe storage (SAFSTOR), the nuclear facility is placed and maintained in

a condition that allows it to be safely stored and subsequently

decontaminated to a level permitting release of the property by NRC.

* In entombment (ENTOMB), radioactive materials are encased in a

structurally long-lived material such as concrete. The entombed structure

is appropriately maintained and surveillance is continued until the

radioactivity decays to a level permitting release of the property by NRC.

In general, only DECON is acceptable for non-power reactors, although for the

cases discussed in 10 CFR 50.82(b)(1)(iii), such as unavailability of waste disposal

capacity, SAFSTOR and ENTOMB are possible options. If either the SAFSTOR

or the ENTOMB decommissioning method is selected, the preliminary DP should

contain (1) the details for preparing the facility for safe storage or for entombment,

(2) plans for monitoring and'surveillance during the storage period, (3) plans to

ensure that funds and other resources will be available for maintaining the facility

NUREG-1S37. PART I 17-4 REv 0.2/96

NUREG-1537, PART I 174 REV 0, 2/96
DEMcoMMssioNING AND PosSEssIO-ONLY AMENDmNT

and completing decommissioning, including the means' of adjusting cost estimates

and associated funding levels over the safe storage or entombment period, and

(4) a comnittment to 'submit an 'updated final DP to obtain authorization before

final decommissioning activities are started.

17.1.4 Release Cnrteria and Final Survey -

To terminate a license, NRC shall determine that the health and safety of the public

will continue to be protected after the facility and site are released. To make such

a determination, NRC should have evidence that radiation'levels at the facility

permit release.

The criteria to release non-power reactor facilities for unrestricted use are as

follows:

(1) (a) no more than 5 microrem (pR) per hour above background at.

'1 meter from the surface measured for indoor gamma radiation

fields from concrete, components, and structures, or

(b) no more than IO milliem (mrem) per year for gamma eitters .-

above background absorbed dose to any person, considering'

reasonable occupancy and proximity (NRC letters dated March 17,

1981, and April 21, 1982)

(2) residual surface contamination consistent with Regulatory Guide 1.86.

Criterion ib,'which considers reasonable occupancy and proxmty, maylallow ;

release of concrete, components and structures that have radiation fields above5;

pR per hour. -However, the staff will consider this release criterio'nonly if the'

licensee can justify its use-for example, if the licensee does not want to

completely demolish a structure and if meetiri the 5-pR-per-hour criterion would

require removing concrete from the structure to the point ofjeopardizing its

integrity., If the area can be secured from occupancy so that criterion lb can be

met, the NRC staff would consider allowing the use of the I 0-mrem-per-year

criterion.

If there is a question of soil, ground water, or surface water contamination at the

site, or if conditions warrant the development of a pathways analysis, the licensee

should contact the NRC staff to discuss release limits (see 'NRC Policy and

Guidance Directive FC 83-23, and 40 CFR Part 141, "National Primary Drinking

Water Standards"). - -

REV. 0,2/96 .17.5 - STANDARD FORMAT AND CoNTENr
CHAPTERI17 - -I

The Commission is considering changes to the release criteria for decommissioning

(see 59 FR 43200). Licensees should ensure that the release criteria used in their

DPs are current.

The decommissioning rule requires the licensee to submit a final radiation survey

plan that shall be described in the DP. Plans for the final radiation survey should

show with a high degree of assurance (e.g., 95-percent confidence) that residual

radioactive contamination levels will meet the release criteria. The procedures,

results, and interpretations described in the final radiation survey plan should be

verifiable by the NRC staff.

17.1.5 Format and Content of Decommissioning Plan

A proposed format for the DP is shown in Appendix 17.1, which lays out step by

step the information required. The format and content are based on two

conditions: (1) the DECON alternatives being used for decommissioning (this

format and content may also apply to the decontamination and dismantling portion

of SAFSTOR) and (2) all fuel has been removed from the reactor and the site. If a

particular project does not meet these conditions, the licensee should contact the

NRC project manager for additional guidance.

The licensee should number the DP sections to follow the format in Appendix

17.1. Each section gives the key items that should be addressed in any DP to

allow timely approval by NRC. Additional guidance for DPs is presented in

American National Standards Institute/American Nuclear Society (ANSI/ANS)

15.10-1981; International Atomic Energy Agency, 1986; NRC draft Regulatory

Guide DG-1005; and NRC 'Guidance and Discussion of Requirements for an

Application To Terminate a Non-Power Reactor Facility Operaig License."

17.2 Possession-Only License Amendment

Possession-only license amendments'are issued to remove the authorization to

operate the reactor from the facility license. NRC generally issues a possessiononly license amendment to a non-power reactor facility after the fuel is

permanently removed from the core and shipped off site. No regulations require

the issuance of a possession-only license amendment. Licensees of non-power

reactors can go directly from an operating icense to an approved DP;

Before the new decommissioning regulations were promulgated in 1988 (53 FR

24018), possession-only license amendments were issued for the remaining term of

the license. Without a regulatory requirement to complete decommissioning

immediately, some licenses with possession-only amendments were renewed at

least once. The new decommissioning regulations did not specifically address

NUREG-1537,PART 1 17-6 REV. 0,2/96 NUREG-1537, PART I 17-6 REV. O. 2f96
DEcoMonSsiomwm AND POSSESSION-ONLY AMENDMENTS

possession-only license amendments, but 10 CFR 50.82(a), as amended, requires

that 'an application for authorization to decommission be submitted within 2 years

after reactor operations permanently cease. Possession-only license amendments

issued by the NRC staff before July 27, 1988, have been allowed to remain in

effect for the term of the license. -E because of factors similar to those in 10 CFR

50.82(bXl)Ciii), the licensee can show good cause for delaying the start ofdecommissioning, the staff will consider renewing these possession-only licenses'.".

issued before July 27, 1988. : -

The purposes of the possession-only license amendment include the following:

* to remove authority from the license to operate the facility

* to continue the licensee's authorization to possess byproduct and special

nuclear material (SNM)

to provide limitations and controls over the radioactive material in order to

protect the health and safety of the public and the facility staff"

* to continue the licensee's authorization to possess the reactor systems

* to allow time to develop, submit, and approve detailed plans for finid'

dismantlement and decontamination of the facility

At the same time that a possession-only license amendment is requested, or at any'

time after it is issued, the licensee may request changes to reduce the technical'.

specifications and other license requirements from those applicable to an operating

reactor. These changes are not required as part of a possessio1wonly license'

amendment application. However, the technical specifications, as written, remain

in force during the possession-only license amendment period.

17.2.1 An Application for a Possession-Only License Amendment

The application for a possession-only license amendment should be made in

sufficient time for NRC review and final action before tlhe projected termination of

reactor operation. If the reactor has been' shut down unexpectedly because of

factors beyond the licensee's control, the application should describe the

circumstances. Factors beyond the licensee's control could include, for example,

abrupt budgetary constraints or equipment failures.

The application for a possession-only license amendment should contain an

explanation of why an amendment is being requested A specific duration for the

possession-only license amendment should be requested, and the activities to be

accomplished and their schedule while the amendment is in effect should be"

R~~v.~ 0_16'-77 w.'~ OMT~Cxr

REV. O. 2/96 . -17-7 .. . - --STANDARD FoRmAT AND Co,\TENT
MCHArm 17

discussed. The licensee should commit to a date to submit the applications for

authorization'to decommission and for license termination. Any factors beyond 2

the licensee's control that affect the schedules should be discussed in detail.

The application for the possession-only license amendment may consist of a

request to delete the authority to operate the facility from the license and a safety

analysis of the permanently shutdown reactor. The licensee may also discuss

proposed license conditions, proposed technical specifications, and proposed

changes to physical security and emergency plans and related safety analyses if it

chooses to change'these aspects of the license. In the safety analysis the licensee

should describe and analyze the shutdown facility and the provisions to control

reactor-related radioactivity to protect the health and safety of the public.

17.2.1.1 Facility License

Proposed changes to the facility license should be consistent with the possessiononly status. The authority to operate the facility should be' removed from the

facility license, and either the authorized maximum thermal operating power of the

facility and the allowable pulse insertion, if applicable, should be reduced to zero

or the paragraph concerning power level should be removed from the license. If

fuel has been removed from the site, authorization to possesi the reactor fuel

should be removed from the paragraph authorizing the possession and use of

SNM. The authority to possess any other SNM or byproduct material that is

specifically authorized by the facility license and that has been removed from the

site or transferred to another license of the decommissioning licensee or to another

licensee should be removed from the facility license. The expiration date of the

license should not be changed.

17.2.1.2 Technical Specifications

The licensee may choose to amend the technical specifications at this time. If so,

the proposed technical specifications should be based on conditions analyzed in the

possession-only safety analysis. It is desirable to remove the fuel from the site

under the operating license, but it is not always possible. In the latter case, the

technical specifications may be more complex, reflecting the need to prevent

criticality and to ensure the integrity of fuel cladding.

Because the reactor will no longer operate, the safety limits and limiting safety

system settings may be removed from the technical specification.' The limiting

conditions for operation may be modified to reflect the fact that the reactor will

not operate again.

Although each situation is unique, technical specification concerning reactor core

parameters, reactor control and safety systems, coolant systems, and experiments

NUREG-1537, PART I 17.8 Rv 0,2/96

NUREG-1537, PART I 17.8 REV 0, 2/96
DECOMMuSSIOtwNG AND POSSESSION-ONLY AmENDmAn

can usually be relaxed. If fuel remains on site, techincal specifications concerning

reactor control and safety systems and coolant system may remain'. Corrosion or

degradation of equipment that confines radioactivity should be monitored. To

limit corrosion, potentially radioactive liquids should be monitored for acceptable- '

pH and electrical conductivity. The licensee should consider the radiological

condition of the facility and protection of the health and safety of the facility staff

and the public when modifying technical specifications for the containment or the

confinement, ventilation systems, emergency power, and radiation monitoring

systems and effluents. The licensee should consider physical barriers or other

constraints necessary to limit personnel access to radioactive and contaminated

areas and should list required radiation monitoring instnjmentation.

The technical specifications should contain a surveillance section that should

include surveillance provisions for each limitation on possession. The surveillance'

specifications should state what is monitored, how, with what'instruments, and on

what schedule, and whit limits are acceptableThe technical specifications should contain a section for describing the site and

facility-, it should contain a concise description of the boundaries of the restricted

area to which the possession-only license amendment will apply. If fuel remains in

the'faility, the reactor coolant system, fuel, and fissionable-material storage

should be described.

The technical specifications should contain a section describing the facility.

administration. Sections on reactor operators may be removed from the'technical

specifications if all fuel is off site. The section on experiment review and approval

may be removed (unless some characterization activities will be considered

experiments). This section should give the information necessary to ensure

continued management of the facility and should describe personze and programs

for specified surveillaince and maintenance activities.; This section will probably be

the least changed of those in the operating license technical speication:

17.213 Emergency, Physical Security, and Operator Requalification Plans

Amending an operating license to a'possession-only license may allow major

changes in emergency and physical security plans if the fuel is removed from the;.

site. These plans are not automatically eliminated. As discussed in Sections 6 and

7 of Appendix 17.1, the licensee may either applyfor a specific exemptiou tom the

requirement to have a physical security and emergency plan or Else submit

amended plans in accordance with 10 CFR 5O.54(p),- (q), and (r). -These plans

should be m6dified 'to' reflectthe facility status and ensure that the health and''

safety of the public continue tobe protected.

REV 0,2/96 - - 17-9-- - STAND' FORMAT AND CoNENT
If the fuel has been removed fro the site, licensed reactor operators are no' longer

needed. Thierefore, the requirements for the operator requalification' program are

no longer needed and the licensee may ask to 'eliminate the plan. A licensed senior

operator'or licensed fuel handler whose requalification is current is required if fuel

is to be moved in the facility.

17.2.1.4 Possession-Only License Amendment Safety Analysis

The purpose of the possession-only license amendment safety analysis is to show

that the facility can be possessed in a way that protects the health and safety of

workers, the public, and the environment. Even though the potential radiological

hazards to the workers and the public are expected to be fewer for a shutdown

non-power reactor, the licensee should consider them carefully. These hazards

should be discussed and analyzed in a possession-only license amendment safety

analysis. The extent of detail in the analysis should depend on how much the

licensee proposes to modify the license and technical specifications.' The safety

analysis for a request to remove only permission to operate the facility may be very

brief because all other requirements will remain.

When the license and technical specifications are to be further modified, the safety

analysis will be more complex. Requirements that are proposed to be removed and

relaxed in the license and technical specifications should be described and justified.

The safety analysis should also form the basis for specifying controlling devices

and administrative procedures in the technical specifications.

Dismantling a reactor is not permitted by a possession-only license amendment.

However, to develop the DP, the licensee may do some limited work to

characterize the facility. Characterization activities may be'descibed in the

possession-only safety analysis and limited by the technical specification. The

reactor facility described in the possession-only safety analysis should be similar to

the operating reactor facility, and the SAR should be referred to as much as

possible. However, some components, instruments, and systems need not remain

operable in the facility. The possession-only safety analysis should state which of

the systems covered by the SAR will remain operable and which will be modified

or deactivated.

The spectrum of accident scenarios for a possession-only safety analysis should be

limited to ones credible for a shutdown reactor and for the residual radioactive

material that are not desc'nibed in the SAR for the operating facility. Postulated

accidents should not subject the'public or the workers to undue radiological

exposure and shall not exceed the applicable regulatory requirements (see

Chapter 13 of this format and content guide).

NIJREG-1537, PART! 17-10 REV 0,2/96

NUREG- 1537, PART I 17-10 REV 0, 2196
DECOMMISSONTNC AND POSSESSION-ONLY AwNDMENrs

Any proposed changes in staff characteristics should be described and justified in

the possession-only safety analysis. Other proposed administrative changes should

also be described and justified.

17±1.5 Changes to Facility Without License Amendment

Because 10 CFR 50.59 applies only to changes under a license authorizing

operation, the licensee should propose a method for maldng operational or

procedural changes under the possession-only license amendment. This should be

similar to the method discussed in Section 9 of Appendix 17.1).

REv. 0,2196 17-11 STANa'X, FoT AND CONTENT

REV . 0, 2196 17-l11 STANDARD FORMAT AND CoNMTE-
Appendix 17.1

Format and Content of Decommissioning Plan

for Non-Power Reactors

STANDARD FOP MT AMD CON70- T
Format and Content of Decommissioning Plan

for Non-Power Reactors

1 SUMMARY OF PLAN

1.1 Introduction

In this section of the decommissioning plan (DP), the licensee should state the type

of reactor, its geographical location, its licensed power level, the term of licensed

operation, the reason for and objectives of the decommissioning, the entity (facility

staff or a contractor) writing the DP, and the organization that will implement the

DP, including the final radiological survey. A synopsis of the DP should follow.

1.2 Background

In this section, the licensee should describe the reactor facility, including the

licensed restricted area, the facility site, and the type, size, and other uses of the

facility buildings. This section should also contain background information about

the owner-operator and a brief operating history.

1.2.1 Reactor Decommissionng Overiew.

The licensee should give an overview of the reactor deco'miussioning process,

including the method selected to decommission, the overall radiological status of

the faclity, the major dimanement tasks to be accomplished, and schedules,.

includi"g the estimated dates for starting and completing the decommissioning.

This overview should briefly discuss the final radiation survey plar'and the..

predicted collective dose equivalent to complete the decommissioninlg. Details.

about these topics'should be given in the appropriate sections'ofthe DP..

1.2.2 Estimated Cost

The licensee should present a cost estimate by principal task for the.,

decomnmuissioning. This estimate should be based on conditions at the facility at the

time the DP is submitted. The estimate should include the costs of shipping and

disposing of waste and of the final radiation survey. In accordance with 1O.CFR

50.75, the licensee shall already have submitted a decommissioning report'that

contains a cost estimate with means of adjusting the cost estimate over the life of

the facility. The cost estimate in this section of the DP should contain much

greater detail than the earlier estimates.

REV. O. 2/96 . I - , ' ' . STANDARD FORMAT AND CONTENT
APPENDIx 17.1

1.2.3 Availability of Funds

The licensee should show how sufficient funds will be made available to

accomplish decommissioning on the proposed schedule. In accordance with 10

CFR 50.75, the licensee shall have already submitted a decommissioning report

that proposed a method or methods acceptable to the Commission to provide

funds for decommissioning.

1.2.4 Program Quality Assurance

The licensee should describe the organizational structure established to ensure that

quality assurance (QA) asues are applied to the planning disenanlement,

radiological surveys, and material shipments. The relationships of the QA function

to the dismantlement organization and to facility management should be made

clear.

2 DECOMMISSIONING ACTIVITIES

2.1 Decommissioning Alternative

In accordance with 1O CFR 50.82(bXl), the licensee shall discuss the approach

chosen for decommissioning the reactor facility (DECON, SAFSTOR, or

ENTOMB) and a description of the activities involved.

The regulations in 1O CFR 50.82(b)(ii) and (b)(1)(iii) generally require that nonpower reactor decommissioning activities be completed without significant delay

and that the health and safety of the public be protected. If the licensee believes

that significant factors such as those specified in the regulations, that are beyond its

control prevent DECON, the licensee should contact NRC before developing a DP

using another method.

In this section of the DP, the licensee should address the elements of the DECON

alternative, summarizing the objectives, the possible fixture use of the site, and any

special factors that relate to the alternative, such as availability of both licensed and

unrestricted waste disposal facilities.

2.2 Facility Radiological Status

2.2.1 Facility Operating History

The licensee should give a detailed and quantitative discussion of the operational

history of the reactor facility that could affect decommissioning planning and safety

'KJ

NUREG- 1537. PART I 2 REV. 0, 2/96
DEcOMMiOmNG AND) PossusioN-ONLY AizMENDS

and from which radiological information can be derived. The licensee should

discuss both routine events and such accidents as radioactivity spills or releases

that significantly contributed to facility radioactivity and contamination levels.

Experience in decommissioning non-power reactors has shown that most

radioactivity to be removed has resulted from routine neutron irdiation of

components. Locations of systems and components that may cause high levels of

radiation and areas of the site that may contain radioactive hot spots should be

identified. This information should be obtained from facility records and personnel

familiar with the facility. As requested in Section 2.2.2 of this appendix, the

licensee should use the operational history along with surveys and sampling to

estimate the facility radionuclide status.

2.2.2 Current Radiological Status of the Facility

The licensee should characeize the reactor-related radioactivity at the f iy.

The information should be based on reactor operating history, calculations,

surveys, and sampling. The radiological characterization is very important because

it will be used in planning tasks, estimating doses, characterizing the radioactive

waste, and estimating its volume. Suggested information includes the following:

* lists of principal radioactive components

* 'lists of isotopes with gamma-ray energy groups, curie strength,'and

geometry

* locations of radioactive sources on reactor and plant layout drawings

* 'discussion of methods and models used to assess the radioactivity'

* cleanup or decontamination tasks already completed

The models and parameters used to calculate the source magnitudes should be

given for all sources, including neutron activation byproduct sources. Predicted

radiation exposure levels frcm systems, structures, and components should be''

evaluated at the time the DP is submitted. 'Complete infonriation may not be'

available for detailed planning. 'However, to the extent practicable, the plan should

describe the sources of radiation on which radiation protection and control are

based. Information in this section'should be-updated as additional radiation

surveys are made. These'updates need not be transmitted to NRC unless required

by Section 9 of the DP. -'

REV. O. 2196 - 3 vSTANDARD FORMAT Am COMENT g
APIENDvK 17.1

2.23 Release Criteria

The licensee should state which release criteria will be applied to the

decommissioning project. Considering the radiological status of the facility, along

with the release criteria, will allow the licensee to plan the decommissioning

activities and tasks.

2.3 Decommissioning Tasks

23.1 Activities and Tasks

The licensee should describe in detail the activities and tasks necessary for the two

phases: (1) preparing the facility for decommissioning and (2) dismantling and

decontaminating the facility. Work should be planned with accurate

determinations of the locations and types of radioactive material to be removed.

On the basis of information in Section 2.2 of the DP, the licensee should describe

the methods, techniques, and equipment, necessary to segment or otherwise

dismantle components and systems. 'All tasks associated with decontaminating the

facility and packaging materials for shipment should be discussed.

The licensee should describe procedures for accomplishing major activities. Health

and safety considerations should be addressed for each task, as appropriate.

2.3.2 Schedule

The licensee should give the sequential schedules for the activites necessary to

plan and complete the facility decommissioning, including the final facility radiation

survey.

For major activities, the relationship between activities and tasks should be shown.

An activity may comprise several specific tasks. Where pertinent, the schedules

for accomplishing interrelated activities and tasks should be delineated. Schedules

and critical path diagrams should indicate the estimated time and resources

required by major activities, including health physics or radiation protection

activities, for the proposed dismantlement and decontamination of the facility. The

proposed schedule should identify realistic locations for disposing of radioactive

waste. The projected date for submitting the final termination radiation survey

report to NRC should be given. The licensee is strongly advised to compare initial

estimates with schedules documented for decommissioning similar non-power

reactors. The licensee should show that the decommissioning can be completed

without significant delay. However, the plan should be flexible enough to allow

for changes in schedules.

NUREG- 1 537, PART I 4 REV 0, 2/96
DECOMaMISSONINo AND PossEsSON-ONLY AmiEmmS

2.4 Decommissioning Organization and Responsibilities

The licensee should describe the organization that will plan and implement the

overall decommissioning, especially management positions. The lines of authority

and the management roles of the personnel protection manager and the faility

safety committee should be shown.

Key positions in the decommissioning organization should be listed and their

functions described. The licensee should diagram the lines of authority from upper

management to the workers. The qualifications and experience of the director of :

decommissioning should be addressed. The staff position with the licensee's onsite

management authority should be stated and the qualifications, duties, and

responsibilities for that position should be described. In addition, the minimum

education, training, and experience requirements should be described for positions

that are important to safety.

The licensee should discuss who manages the contractors and what'jobs

contractors do. Contractor qualification requirements should be defined (see

Section 2.6 of this appendix). The discussion should show tha the licensee

continues to be responsible for overall supervision,'compliance with applicable'

regulations, and protection of the health and safety of thepublic.

The management policy and organizational structure related to aesring that

occupational radiation exposures are as low as is reasonably achievable (ALARA)

should be described. The responsibilities and the activities of management and

health physics personnel responsible for radiation protection and the ALARA;

program should be discussed.

The role and composition of the facility safety committee should e described,

especially its review and auditing authority. Attention should be given to its

experience with radioactive waste and radiological exposure.

2.5 Training Program

Dismantlement and decontamination may be a new experience for the reactor staff;

therefore, special training may be required. Also, contractors should receive :

training'on the DP and the site. The training plans should demonstrate that the

licensee is aware of the differences between normal operation and

decommissioning. - : -

This section should describe the proposed training program. The training should

cover the applicable requirements of the DP, principles and techniques of

decontamination and dismantlement activities, industrial hygiene, health physics,

REV. 0, 2.6 5 STANDARD FoiM.T CONTENT

RF\v.0, 2/96 . s STANDARD FORMAT AND CoNIENT -
AmzENDDC 17.1 - ---

and use and maintenance of monitoring and safety equipment. Compliancewith

10 CFR Part 19 should be discussed as required to accomplishDP activities.

The qualifications, experience, and responsibilities of persons responsible for

trining should be discussed. The following should be discussed for all training

activities: training status for all personnel (including contractor personnel),

training required for new employees, frequency of refresher training, and personnel

certifications.

2.6 Contractor Assistance

The licensee may use contractors to perform some or all of the decommissioning

activities and tasks. However, the responsibility for health and safety and

compliance with the regulations during all aspects of decommissioning rests with

the licensee. The licensee should ensure that contractors comply with all

applicable license conditions and regulations.

it is important to discuss the administrative control system used by the licensee to

ensure adequate health and safety protection. It is also important to discuss how

quality assurance will be achieved. For each contractor, the plan should describe

the type and scope of work to be accomplished and the relationship of the

contracted work to the schedule for the other activities. The level of qualifications

and experience required of contractors and their employees should be discussed.

The prior history and performance of the contractor on non-power reactor

decommissioning projects should be discussed.

2.7 Decontamination and Decommissioning Documents

and Guides

The licensee should discuss the health physics and industrial health criteria and

standards that will guide the activities described in the DP. Relevant documents

include Regulatory Guide 1.86, ANSIIANS 15.10-1981, and Occupational Safety

and Health Administration (OSHA) requirements.

Much information has been published about decommissioning. Some documents

are the following: Manion and LaGuardia, 1980; NRC draft Regulatory Guide

DG-1006; NRC lE Circular 81-07; NRC Information Notices 83-05 and 85-92;

NRC letters dated March 17, 1981, and April 21, 1982; and NUREG/CR-1756

and addendum. The licensee should review the literature on the decommissioning

of similar non-power reactors, including DPs that have been submitted to and

approved by NRC. The licensee is advised to use all possible information in

preparing the DP.

NUREG-1537, PART I 6 REV. 0,2/96

NUREG-1537, PART I 6 REV. 0. 2/96
DECOMMISSIONING ANDPOSSESSION-ONLY AIm ENS

The licensee shall also show compliance with the regulations, including 10 CFR

Parts 20,30, 50, 70,71, and 73 as they apply to decommissioning and with

49 CFR Parts 170 through 189.

3 PROTECTION OF THE HEALTH AND SAFETY OF

RADIATION WORKERS AND THIE PUBLIC

3.1 Radiation Protection

3.1.1 Ensuring As Low As Reasonably Achievable Radiation -

Exposures

The licensee should state its policy for Imaintaining radiation eexposures'and''

releases ALARA during decommissioning and describe howthe policy will be

implemented throughout the decommissioning process. The DP should 'descrbe

either a modification of the edstiig fcility ALARA'program or the creation of a

complete ALARA program specifically for decommissioning.

Management positions responsible for ensuring radiation'protection and

implementation of the ALARA program during decommissioning should be

described. The licensee should show where the manager of the ALARA program

and the radiation protection program fit into the overall management structure and

should explain how authority to implement the program is derived and applied.

Information rox previous decommissioning reports and availe liierature may be

used to assist in developing the program.

3.1.2 Health Physics Program''

The licensee should describe in detail the health physics program for

decommissioning activities. The overall management and the'authority,

responsibility, and duties of each position should be identified. This section should''

include a diagram showing the relationship of the health physics manager to the

decommissioning program manager and should discuss the authority to stop

potentially unsafe practices.

The licensee should describe radiation protection methods for the public the'

decomrniissioning Staf, and radiation workers. These methods'should include

protection from radiation fields, contamination, and airborne'radiation hazards

created during the' deconissioning process. This section should contaiini'

information about survey and personnel m'orutoring' equipment, including the

criteria for selecting the equipment and requirements fdr'miaintaining, storing, and

REV ,216 7- STNDAD FRMATAN! COTEN

REV 0, 2/96 7 . -, STAwDARD FoRmAT Am Comvrr
inAPPEMD 17 1

calibrating it. The policy, frequency, and procedures for conducting radiation

surveys, monitoring effluents and personnel and assessing and documenting

radiation exposures should be described in detail.

The radiation protection practices used by the licensee to meet 10 CFR Part 20

requirements and the procedures for implementing the ALARA program should be

described in detail. The licensee should clarify whether the existing health physics

program will continue during decommissioning or will be modified.

3.1.3 Dose Estimates

For each major task in which radiation is a factor, the licensee should estimate the

total cumulative and individual maximum dose equivalents for radiation workers to

complete the task. Both external and internal exposures should be estimated. In

addition, this estimate should include credible potential exposure pathways and

doses to the public. Methods to ensure compliance with 10 CFR Part 20 and the

ALARA program should be explained The licensee may gain insight into dose

estimates by reviewing final reports from non-power reactors that have been

decommissioned.

3.2 Radioactive Waste Management

3.2.1 Fuel Removal

New and irradiated fuel at non-power reactors should be shipped off site before

reactor dismantlement or decommissioning begins. Fuel may be removed from the

reactor and shipped under the existing operating license. The DPineed only give

the date of shipment and disposition of the fuel. The licensee should state that fuel

removal and shipment are not included in the DP. The Department of Energy

(DOE) has agreed to accept all fuel from licensed non-power reactors under the

Nuclear Waste Policy Act of 1982. Fuel must be shipped under applicable NRC

and Department of Transportation regulations. The DOE will help the licensee

plan for and implement fuel shipments.

The regulations do not require that fuel be completely removed before

decommissioning is authorized but the presence of fuel on site during

decommissioning may constrain planning and work flexibility and limit the amount

of work that can be done. For example, in planning the decommissioning, the

licensee must analyze any possible accidents that could occur if demolition took

place in"the same building as the fuel. The emergency plan and security plan must

continue to be in effect if the fuel is on site. A licensee considering the inclusion of

fuel disposal in the DP should contact the NRC staff

NUREG-1537, PART I 8 REv 0.2/96

NUREG- 1537. PRT I 8 REV 0, 2196
DECOMMISSiONING AND PossEsSIoN-ONLY AMDENTS

3.2.2 Radioactive Waste Processing

The licensee should describe the systems and procedures used to detect, control,'

and process radioactive wastes before disposaL 'These may be gaseous, liquid, or'

solid radioactive wastes generated during decontamination and other

decommissioning activities. The discussion should address radioactivity levels,

volumes of radioactive waste, exsting treatment and processing systems that will

be retained, process control program, disposal schedules, and radioactive waste

control systems to comply with the regulatory requirements and the guidelines of

the ALARA program.

3.23 Radioactive Waste Disposal -

The licensee should give the quantities and types of the radioactive materials for

disposal 'uring decommissioning based on the radiological status of the facility as

discussed in Section 2.2.2 of the DP that result from the decommissioning

activities. It should explain where and how all materials will be disposed of and

how the materials will be transported to disposal sites. It should cite applicable

regulations and license conditions and explain how compliance will be achieved.

The licensee'should discuss radiation measurements, survey methods, and the

criteria for determining which materials should be disposed of in licensed locations

and which materials may be disposed of in unrestricted sites. The licensee should

discuss regulations and guidance for determining what is radioactive waste (NRC

1E Circular 81-07 and NRC lE Information Notices 83-05 and 85-92). It should

also discuss the experience and qualifications of any persons who will handle or

control radioactive materials in unrestricted areas, as during tran

3.2.4 General Industrial Safety Program

The licensee should discuss the protection of personnel from potentially hazardous

noniadiation exposures and situations resulting from the decommissioning. These

haizrds might include conditions associated with controlled demolition; open

holes; nonradioactive airborne debris;'nonra'dioactive airborne hazards generated

by burning or cutting; including displacement of oxygen; and solvents used to

remove radioactive contamination. Many ofthesebiazards maybe new 'to the

reactor operating organization. The licensee should describe the industrial safety

program and show that all applicable OSHA and industrial safety requirements will

be met. The authority and responsibility of each nonradiologickal safety positiorn in

the decommissioning organization should be stated. In addition, this section

should specify the criteria for selecting equipment and methods' to control

nonradioactive exposures and hazards. Nonradiological accident prevention and

response should be discussed.

REv. 0,2/96 9 STANDARD FORMAT CoNT

REX'. O. 2/9 9 STAmARD FoRmAT AND CoNmw
APPENDIX 17.1

3.3 Radiological Accident Analyses

The licensee should discuss potential radiological accidents that could affect the

public or occupational health and safety and present its conclusion as to the

acceptability of the results of the accident analysis. Accidents that are directly

related to decommissioning and that differ from accidents related to normal reactor

operation or maintenance should be emphasized. Sufficient detail should be

included to clearly define and analyze the consequences of any significant potential

accidents. If NRC has approved fuel removal as a part of the DP, radiological

accidents related to fuel storage during decommissioning should be discussed.

All potential releases of radioactivity to or from controlled areas should be

discussed.

4 PROPOSED FINAL RADIATION SURVEY PLAN

The final radiation survey report is very important because it forms the basis for

verifying that the facility, site, and environs meet radioactivity levels that permit

NRC release of the property.

In this section of the DP, the licensee should propose a plan for conducting final

radiation surveys. The plan should describe how the facility and site will be

surveyed to demonstrate compliance with the release criteria. The description

should include the following information:

• methods to ensure that sufficient data about peinent structures, systems,

components, equipment, the site, and the environs are included in the survey

(including maps, diagrams, and plant layout drawings) %

* types, calibrations, and operating conditions of instruments to be used

* methods to obtain and analyze data, including the methodology selected to

translate instrument readings or sample analysis results into appropriate units

for the report (e.g., dpml100 cm2, Ci/g of soil, curies, PMh)

* comparisons with preoperational radiation survey results and other data on

background radiation

* methods for auditing and verifying data

* quantitative error analyses of the results

'KJ

NtJREG-1537. PART I 10 REV 0.2/96

NUREG-1537,PARTI1 10 REV 0, 2/96
DECOMMISSIONIDG AND POSSESSION-ONLY AMENDME

* . basis and methods for making statistical inferences from the data selected to

ensure that all significant residual sources of radiation are found and

quantified

The discussion of the radiation survey plan should include an outline of the final'

report which will contain the results of the final survey and cite relevant references.

NUREGICR-2082 contains significant information about final survey monitoring.

The final survey report should be submitted when the licensee requests theNRC

confirmatory termination survey. The report should sunmarize the

decommissioning activities conducted at the facility to prepare for license.

termination. It should show how actual quantities of radioactive material waste

removed from the site, and cumulative and individual exposures compare with the

estimates of the DP. The report should also discuss any unexpected results or

findings that differ from those predicted by the DP, the resolution of these issues,

and any significant insights to assist other licensees that will decommission their

facilities one day. The report should present the results of the final radiation and

environmental surveys, and their respective analyses, and should make direct

comparisons to radiological guidelines and limits for release of the facility. The

report should discuss what was done with radioactive material that was not

disposed of as waste, such as fuel, sources, and components that were transferred

for use or possession under other licenses.

The radioactivity at the time of decommissioning will consist of contamination

(spilled, deposited, and adherent) and radioactivity formed within components by

neutron irradiation .AII sources of radiation should be considered and either

removed or reduced sufficiently for release of the facility. The licensee should

discuss how compliance with the release criteria was achievedk These criteria are

discussed in Section 2.2.3 (above).

If there is a question of soil, ground water, or surface water contamination at the

site or if conditions warrant the development of a pathways analysis, the licensee

should contact the NRC staff to discuss release limits.

Currently, no radioactive material, as determined by acceptable industry standards,

may be disposed of at unlicensed locations, such as public landfills, without prior

NRC approvai (see NRC IE Circular 81-07 and NRC Information Notice 83-05:

and 85-92).

5 TECHNICAL SPECIFICATIONS

After the nuclear fuel is removed from the reactor and shipped off site, most of the

technical specifications for the operating license will not apply if the license has

been amended to possession-only or modified by an order to decommission.

REV . O. 2/96 I1I STANDARD FORMAT AND CoNIErN
APiENDuc 17.1

During the, decommissioning phase, technical specifications should be derived from

an analysis of the radiological health and safety of personnel and from an.

environmental assessment of the decommissioning of the facility. The format of

the technical specification should follow the appropriate sections of ANSI/ANS

15.1 as discussed in Chapter 14, 'Technical Specifications," of this format and

content guide. They should include required equipment, radiation protection, and

administraive requirements such as those pertaining to organization, review and

audit, procedures, reporting, and records. The analysis should lead to the

conclusion that projected radiation exposures of decommissioning personnel and

the public are ALARA and are small fractions of the respective regulatory limits

(10 CFR Part 20). The decommissioning technical specifications and procedural

controls should contain the safety precautions necessary during the

decommissioning phases. The technical specifications should contain the

following:

* a section imposing limiting decommissioning conditions at the facility

comparable to the limiting conditions for operation for required equipment

and operational conditions

* a section providing for surveillance of the required equipment and conditions

for decommissioning

* a section describing the residual defueled facility and site to which the

decommissioning order will apply

* an adm it section that outlines the management structure, provides

for review and audit functions, provides for development and use of

necessary procedures, and contains reporting and record-repention

requirements. %

These controls should allow margins sufficient for onsite reaction to changing or

unanticipated events.

6 PHYSICAL SECURITY PLAN

This section should contain a description and schedule of proposed changes to the

NRC-approved physical security and material control and accountability plans,

when applicable Information in this section that must be protected from public

disclosure under 10 CFR 73 21 or 10 CFR 2.790 shall be submitted under separate

cover and identified appropriately.

If the fuel is shipped off site before or at the time decommissioning activities are

authorized, the licensee may request relief from the physical security plan under

10 CFR 50.54(p). The requirements for this plan depend on the amount and

NUREG-1537.PARTI 12 REV 0, 2/96
DECOmNO.SSIONTNG AMD PossEsSIoN-ONLY Awmlrs

enrichment of special nuclear material (SNW on site.'- Removal of fuel may allow

the level of site security required by the regulations to be reduced. The licensee

will be required to take measures to control public and worker access to radiation

areas. Such measures could be included in the revised technical specifications. ra. . 0

7 EMERGENCY PLAN

The regulations require a facility to have an emergency plan as long as it has a

license. However, if the fuel, which is the major sou rce of radioactive material on

site, is removed, accidents that involve enough radioactive material tb reach the

emergency action levels spelled out in the plan may be impossible. Fuel removal

may allow a major reduction in the level of emergency preparedness 'required at the

facility. Under 10 CFR 50.54(q), the licensee, without prioi NRC approval, may

make changes to the emergency plan that do not reduce its effectiveness. Changes

that reduce the effectivene'ss of the plan'cannot be made under 10 CFR 50.54(q)

and shall be approved by NRC before the changes are put into'effect. To

completely eliminate the requirement for the emergency plan, the licensee should

apply for a specific exemption and address all the factors in 10 CFR 50.12.

8 ENVIRONMENTAL REPORT

In accordance with 10 CFR 51.21, theNRC staff shall prepare an environmental

assessment for the proposed decommissioning of a non-power reactor and the

termination of the facility license. The assessment should be written in compliance

with 10 CFR 51.30 and should be based on information given by the licensee in an

environmental report (ER). The licensee may include the ER as a section of the

DP or submit it asa separate report with the DP. To the extent possible, the

content of the ER should follow 10 CFR 51.45. The ER-should also address the

following:

* the collective dose equivalent to workers for the entire dismantling and

decommissioning project

* potential exposures of the public, including the most exposed person, from

radioactive effluents released during the decommissioning activities and

radioactive waste shipments

* anticipated potential exposures of the public after license termination

9 CHANGES TO THE DECOMMISSIONING PLAN

Because not all events'and discoveries during decomnuissioning are predictable, the

DP may need to be changed. Unless the DP contains a method to make changes

REV. O. 2/96 13 STmDARD FORMAT AND CONTENT
ALAPPENDix 17 1

without prior NRC approval, NRC will have to approve all changes to the plan, no

matter how small or insignificant with regard to safety. Because the DP takes the

place of the facility safety analysis report, the licensee should describe provisions

for changing the approved DP and DP procedures and for conducting tests. The

provisions for changes should be similar to those described in 10 CFR 50.59 for

operating reactors. The DP or decommissioning technical specifications should

describe the following:

* the ldnds of changes the licensee can make without prior NRC approval

[similar to the requirements of 10 CFR 50.59(a)(1)J

* the definition of an unreviewed safety question [similar to the requirements

of 10 CFR 50.59(a)(2)]

* the way records of the changes made will be held by the licensee and the

contents of the records [similar to the requirements of 10 CFR 50.59(bXl)]

* a requirement to submit a report of these changes to NRC [similar to the

requirements of 50.59(b)(2)]

* the length of time records will be held by the licensee [similar to the

requirements of 10 CFR 50.59(b)(3)]

* the procedure for making changes to the DP technical specifications or

changes that involve an unreviewed safety question [similar to the

requirements of 10 CFR 50.59(c)]

Further information about changes to DPs can be found in license Amendment 85

for the Fort St. Vrain facility (Docket No. 50-267), dated Nov&sber 23, 1992.

10 REFERENCES

American National Standards Institu/American Nuclear Society, ANSIT/ANS

15.1, 'The Development of Technical Specifications for Research Reactors," ANS,

LaGrange Park, Illinois, 1990.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15. 10, "Decommissioning of Research Reactors," ANS, LaGrange Park, Illinois,

1981.

Code of Federal Regulations, Title 10, "Energy," and Title 49, "Transportation,"

U.S. Government Printing Office, Washington, D.C., revised periodically.

NUREG-1537. PART I 14 REv 0,2/96

NUREG- 1537. PART I 14 Rrv 0, 2/96
DECOMISSIONwNG AND PossEsIoN-ONLY AmuNDMrrs

Manion, W. J., and T. S. LaGuardia, "Decommissioning Handbook,"

DOEJEVI10128/1, prepared for the U.S. Department of Energy, November 1980.

U.S. Nuclear Regulatory Commission, Draft Regulatory Guide DG-1006,

'Records Important for Decommissioning of Nuclear Reactors," September 1989.

U.S. Nuclear Regulatory Commission, IE Circular. 81-07, 'Control of

Radioactively Contaminated Material," May 14, 1981.

U.S. Nuclear Regulatory Commission, IE Information Notice 83-05, "Obtaining

Approval for Disposing of Very Low-Level Radioactive Waste," February 24,

1983.

U.S. Nuclear Regulatory Commission, IE Information Notice 85-92, "Survey of

Wastes Before Disposal From Nuclear Reactor Facilities," December 2, 1985.

U.S. Nuclear Regulatory Commission, letter from John F. Stolz to Dr. Roland A.

Fbiston, Stanford University, "Use of the 10 mrem per Year Release Criteria,"

March 17, 1981.

U.S. Nuclear Regulatory Commission, letter from James R Miller to Dr. Roland

A. Finston, Stanford University, "Use of the 10 mrem per year Release Criteria,"

April 21, 1982.

U.S. Nuclear Regulatory Commission, NUREG/CR-1756, "Technology, Safety,

and Costs of Decommissioning Reference Nuclear Research and Test Reactors,"

March 1982, and Addendum 1, July 1983.

U.S. Nuclear Regulatory Commission, NUREG/CR-2082 "Monitohing for

Compliance With Decommissioning Termination Survey Criteria," July 1981.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.86, "Termination of

Operating Licenses for Nuclear Reactors." 1974.

REV. 0. 2/96 is Sm2'z1w FORMAT AND CONThN7

REX. O. V%9 15 STANDARD FORMAT AND Co~rE'T
18 E1IGHLY ENRICHED TO LOW-ENRICHED

URANIUM CONVERSIONS

The licensee should use the'guidanceiin'this chapter of the SAR in preparing the

safety analyses to support the conversion from highly enriched uranium (HEU) fuel

to low-enriched uranium (LEU) 'fuel in an eisting non-power reactor.

In 1986, NRC promulgated the final tile 'Limiting the Use of Highly Enriched -

Uranium in Domestically licensed Research and Test Reactors" in Section 50.64

of Part 50 to Title 10 of the Code ofFekralReguZations (10 CFR 50.64, 51 FR'

6514, February 25, 1986). Thisrule requires 'that NRC licensees of non-power

rectors, who are authorized to possess and use HEU fuel, convert from'HEU to

LEU fuel if Federal funding is available'unless a facility is specifically exempted

because of a unique purpose as defined in 10 CFR 50.2. Each affected licensee is

required to submit a conversion proposal to NRC for implementing the regulation.

The conversion proposal and the supporting safety'analyses should consider Il',.

changes to the facility,' license, and procedures necessary to convert'from BEU to

LEU fuel. 'NRC will determine if the proposed changes and supporting safety

analyses ens-ure that the public health and safety will be protected. If NRC finds

that the proposali acceptbly implements the rule, it will issue an enforcement order

directing the conversion and any necessary changes in the license, facility, and

procedures. In most cases, these orders will modify the license under 10 CFR

2.202.

NRC has selected to use enforcement orders to require the conversion from HEU

to LEU fuel, rather than to amend the operating license. This process was selected

so that NRC will be responsible for defending the conversion order. -This process

ensures that any intervention will be made against the NRC orderland not against

the licensee's proposal for conversion. r (Intervention allows a person or persons

whose interest may be affected by the order to request a hearing or to participate

as a party in hearings.)

The licensee should limit its conversion proposal to changes required to convert

from HEU to LEU fuel because the regulations and the associated NRC order are

solely for that purpose. The licensee should show that the proposed changes are

required to make a safe and effective fuel conversion. If other changes are desired,

they should be proposed separately in accordance with the appropriate regulatory

requirements (e.g., 10 CFR '50.90).

Factors beyond the control of the licensee in the conversion process may require

modifications to the facility or its operating characteristics. For example, the

NRC, the U.S. Department of Energy, and the non-power reactor community

agreed that a standardized LEU fuel plate would be designed for plate-type

REv 0.2196 18.1 STANDARD FOIMAT AND UO1fltNT

REV 0, 2/96 18 1 STANDARD FORMAT AND CONTE1,rr
CHAPTER 18

reactors. The characteristics of the approved LEU fuel plate could be sufficiently

different from the HEU fuel plate in use at the facility to require changes to the

reactor. The licensee should note the physical reason for the required changes and

should prepare a safety analysis of the effects of the changes.

The Commission has determined that the conversion to LEU should not result in

either a significant reduction in the nuclear capabilities of the reactor or in an

increase in capability with a significant decrease in safety margin. Therefore,

conversion to LEU fuel should allow non-power reactors to increase or generally

maintain the nuclear capabilities similar to the current HEU cores if there is no

significant change in the safety analysis results. That is, the safety analyses should

demonstrate that acceptance criteria to ensure public health and safety will be met

and that the change is needed to accommodate the design differences for the LEU

core and fuel.

The information provided in the conversion proposal and associated safety

analyses should reflect the current state of the technology, applicable regulatory

guides, and industry codes and standards, as well as developments in nuclear safety

and experience gained in the construction and operation of non-power reactors.

The proposal and safety analyses must account for changes in applicable NRC

regulations which were promulgated after the facility was licensed. The design

information should reflect the most advanced state of design at the time.

Conformance with this standard format is not required, but the content of the

proposal and associated safety analyses should be sufficient for NRC to arrive at

all necessary conclusions to approve the proposal and issue the conversion order.

NRC will compare the content of the proposal and associated safety analyses

against this document and will review it using Chapter 18 of NUAEG-1537,

Part 2. If the proposal and associated safety analyses are not condprehensive, the

staff will request additional information before continuing its review.

Because of the variety of non-power reactors with differing operating

characteristics, this chapter attempts to include all possible areas that may be

affected by the conversion. Therefore, some of the content may not be applicable

for a particular facility. In such cases, the licensee should briefly explain why some

topics are omitted.

Portions of the SAR (e.g., site characteristics) that will not change because of the

conversion need not be repeated. However, although not required for the HEU

conversion to LEU fuel, the licensee may consider submitting a complete revision

of the SAR separate from the conversion proposal to have a single complete

reference that describes the current licensing bases for the LEU facility.

NIJREG-1537. PART I 18-2 REV 0.2/96 NUREG- 1537, PART I 18-2 REV 0. 2/96
HEU TO LEU CONVERSIONS

The numbering system in Appendix 18.1, Format and Content for HEU to LEU

Conversions at Non-Power Reactors," corresponds to the numbering system of the

conversion SAR.

It is organized to be consistent with recent licensee SARs and NRC safety

evaluation reports for non-power reactors with changes related to specific HEU to

LEU conversions. If there is no change to the content of a section for conversion,

the licensee should state that fact and should briefly discuss the reason.

Reference made to a chapter other than this one (Chapter 18) alludes to another

chapter in the format and content guide. Reference to a section alludes to a

section in the appendix.

Additional guidance may also be found in Part 2 of this document as well as in

other chapters of Part 1. Also, the licensee can consult such documents as the

International Atomic Energy Agency (LAEA) documents IAEA-TECDOC-223,

-324, and -643 for additional guidance.

REv. 0, 2/96 18.3 STANDA.P.D FORMAT AND CoItNr

REX'. O. 2196 18-3 STANDARD FORMAT AND CONmENT
Appendix 18.1

Format and Content for HEU to LEU Conversions

at Non-Power Reactors

STANDARD FORMAT AND CoN7ENT
Format and Content for HEU to LEU Conversions.

at Non-Power Reactors

1 GENERAL DESCRIPTION OF THE FACILITY*

1.1 Introduction

In this section the licensee should give a brief overview of the conversion from

BEU to LEU fuel. The overview should summarize the physical, nuclear, and

operational characteristics of the facility for the LEU fuel and should refer to the

changes and associated sections ofthis appendix needed to convert from HEU

fuel. It should also have comparisons and references to the existing reactor

described in the HEU SAR

1.2 Summary and Conclusions of Principal Safety

Considerations

All changes to the facility SAR should be briefly discussed. Emphasis should be

placed on the changes that significantly affect facility safety. Referelce should be

made to the section(s) in which these changes are analyzed in detail. The

referenced SAR section(s) should explicitly mention and compare the changes to

the current HEU SAR.

1.3 Summary of Reactor Facility Changes

A summary description of the changes to the reactor and the site associated with

the conversion to LEU fuel should be provided. The types of HIU fuel currently

used in the reactor and the proposed LEU fuel should be briefly dbscribed. This

summary description should refer to subsequent discussions on detailed fuel design

and should reference documents thaf describe the proposed type of LEU fuel.'

Both the HEU and the LEU fuel 'should be brnefly described and compared. If

NRC has previously reviewed and accepted the proposed fuel, reference should be

made to the acceptance document (e.g., NUREG-1281, -1282, or-1313).

Additionally, this portion of the BEU to LEU conversion SAR should discuss any

other proposed changes to the facility and give reasons why these changes are

required by the conversion.

'The numbering system m thus appendix (Sections I through 15 8) corresponds to the numbering

system in the converson SAR

REV 0,2/96 I STA.iw FORMAT AND CONTENT

REX' 0, 2/96 I STANDARD FoRmAT AND CoNmw
APPENDxIX 18 I

1.4 Summary of Operating License, Technical

Specifications, and Procedural Changes

In this section, the licensee should summarize any required changes to the facility,

operating license, technical specifications, and operating procedures that are

proposed to accompany the conversion and should reference the sections in which

they are discussed. The following changes should be considered:

* possession limits on special nuclear material

* required provisions to possess HEU and LEU fuel simultaneously

* technical specification changes to specify such LEU fuel and reactor

parameters as

- dimensions

- materials

- uranium loading

- enrichments

- core sizes and number of fuel elements

- reactor physics parameters

- process variable values

- safety limits, limiting safety system settings, and limiting conditions

for operation.

1.5 Comparison With Similar Facilities Already

Converted

If possible, the facility should be compared briefly with similar faicilities that have

already been converted from HEU to LEU fiueL The information should include

the following:

* Similarities and differences between the facilities in design, construction,

fuels, uses, and operational characteristics.

* Problems related to conversion that were encountered and resolved at the

other facilities.

* Discussion of how problems are addressed in the plans for conversion of the

subject reactor, if applicable Reference appropriate sections of the SAR

where details are provided.

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- - HEU TO LEU CoNvERsIONS

2 SITE CHARACTERISTICS

Among the site characteristics to be described are geography, demography, nearby

transportation, industrial, and military facilities; meteorology geology-, seismology-;

and hydrology. Generally, there should be no change to these topics with a

conversion from HEU to LEU fuel. However, it is possible'that new information

on site characteristics has become available since the last revision of the SAR. For

example, it is possible that demography and nearby transportation, industrial and

military fidlities around the reactor facility have changed since the last SAR

revision. If these changes are significant to the safety analysis of the reactor or if

the licansee wants the SAR to be current, these new factors should be justified and

submitted to NRC as an update to the HEU SAR preceding the submittal of the

LEU SARI Changes to the site characteristics caused by the passage of time or the

discovery of new information unrelated to the conversion cannot be approved by

NRC using the conversion order. After separate review and approval by NRC,

these changes can be described in this section of the LEU SAR. The existing site

characteristics should be analyzed to determine what changes, if any, have

occurred in the safety analysis of the LEU-fueled reactor as compared to the same

analyses for the lEU-fueled reactor.-

If the changes in site characteristics are not significantly different for the LEU-.

fueled reactor when compared to the HEU-fueled reactor, the bases for this

conclusion should be simply stated.

If the conversion from HEU to LEU changes the effect that nearby industrial or

transportation uses have on the LEU facility, such as ensuring that personnel can

continue to perform their functions in the event of an incident at the industrial or.

transportation facilities, the licensee should describe the changes in the facility.

design and operating criteria to ensure safe facility conditions. Such changes

should be listed in this section, and described and analyzed where the design or

operating criteria of the specific systems, structures, or components are discussed

Most non-power reactors release small quantities of radioactive argon41 to the,-

unrestricted area during normal operations.; Any changes in this and other

radioactive effluents resulting from the proposed conversion and the impact on the

topics of this section should be mentioned. However, comparisons of potential

doses to the public considering current popiulation distributions should be made in

Section 11, as well as comparisons with regulatory requirements.

The site should be shown to remain suitable for the safe operation of the reactor.

Additional general guidance may be found in American Natiodal Standards

lnstitute/Arnerican Nuclear Society (ANSI/ANS) Standard 15.7-1977

REv 0,2196 3 STANDARD tORMAT AND CONTENT

REX' O. 2196 3 STANDARD k'ORMAT AND UOS`TENT
APIV8XISI

3 DESIGN OF STRUCTURES, SYSTEMS, AND

COMPONENTS

The changes to principal architectural and engineering design bases for the

structures, systems, and components of the facility should be discussed here.

Generally, there should be no change to this section as a result of converting a

non-power reactor to LEU fiel, and extensive reference can be made to the

current HEU SAR.

In this section, the licensee should reference all significant standards and guides

applicable to the-conversion, including rules and regulations, NRC staff reports

(NUREGs), regulatory guides, and the ANSIIANS standards used to design all

proposed modifications to the reactor facility. In this section, the licensee should

also describe how the facility is designed to accommodate potential damage from

wind, water, and earthquakes, as well as any changes to the design bases for

structures, systems, and components that are a result of the HEU to LEU fuel

conversion. The discussions should note the HEU SAR design bases and details

for the proposed conversion to LEU fuel to offer reasonable assurance that the

associated structures, systems, and components will perform their design function.

The licensee may be able to verify that the conclusions of the HEU SAR remain

valid by comparing key HEU and LEU characteristics.

The applicable design standards, codes, and criteria that may be affected by the

conversion to LEU fuel should be described. In the case of structures, for

example, building codes and allowable leak rates for the containment or

confinement should ensure that any possible potential increase in fission product

inventory associated with the LEU fuel element inventory in accident scenarios is

acceptably controlled.

If the effective conversion of the reactor from ELEU to LEU fuel requires changes

to the systems, structures, or components that are designed to limit the effect of

environmental or other factors on the facility, the changes should be listed here,

and described and analyzed where those systems, structures, or components are

discussed. The information should provide the design bases and functional

descriptions, concluding that the changes will not significay increase potential

radiological exposures to the facility staff the environment, or the public.

4 REACTOR DESCRIPTION

4.1 Reactor Facility

A sunMay description of the changes to the reactor facility for the conversion to

LEU fuel should be presented in this section. It should include replacement of the

NUREG-1537, PART I 4 REV 0,2/96 NUREG- 1537, PkRT I 4 REV 0, 2/96
HEU TO LEU CONVEySONS

HEU fuel and necessary changes to the balance of the reactor facility to

accommodate the conversion. The design functions of the reactor with LEU fuel

should be demonstrated to perform acceptably over the range of potential

conditions. The similarities and differences between the currently licensed HEUfueled reactor and the proposed LEU-fueled reactor should be tabulated.

In addition to fuel enrichment, other parameters directly affected by the conversion

may be the density offthe uranium, the cladding material and dimensions, the

chemical composition and dimensions of the fuel, irradiation behavior, and

configuration of fuel elements. Although not generally anticipated, any necessary

changes to in-core components, such as the core grid plate and neutron moderator

or reflector, should be described. The licensee should discuss whether any

significant operational characteristics or procedures will be changed for the

proposed LEU-fueled reactor and should provide reference to associated analyses.

4.2 Reactor Core

A detailed discussion of the components and structures in the reactor c6ore should

be provided, including a summary description of the core chinges for the

conversion to LEU fuel. The licensee should submit figures showing the locations

and geometries of the components in the two cores and fists comparng rtant

design parameters and operational characteristics.

4.2.1 Fuel Elements

The LEU fuel elements should be compared to the HEU fuel elements. Any

changes resulting from the lower enrichment and possible higher uranium

concentration in the LEU elements should be included. The licejsee should

discuss in detail the niechanical design of the fuel element, volume ratios of fuel to

moderator and fuel to coolant, uranium burnup, fission product barrier (cladding)

and retention capabilities, and thermal capabilities and characteristics of fueled

components. Dimensions such as water gap thickness and fuel element spacing

also should be given. If applicable, control and dummy elements should be

described.

NRC issued safety evaluations for the types of fuel previously found acceptable in

conversions of non-power reactors from lIEU to LEU fuel (i.e., NUREG-1281, -

1282, and -1313). If one of these fuels is used for the conversion, the SAR should

show that the characteristics of the proposed fuel in the converted reactor will be

consistent with the characteristics reviewed and evaluated in the corresponding

staff safety evaluation. Conformance of fuel fabrication with applicable guidance

in ANSIIANS 15 2-1990 should also be confirmed for plate-type fuels. For a fuel

design that differs from those previously found acceptable in NUREG-128 1,

R.Ev.O,21_ TNADF~v N _ OT

RE%,. 0. 2/96 5 STANDARD FoRmAT AND CoNTENT
APPLYM 18 1

-1282, and -1283, analyses and bases similar in content to those reviewed and

accepted in the NRC documents should be given.

The HBEU and the proposed LEU thermal capabilities and characteristics of fueled

components should be compared, and power density and peaking and all allowed

core configurations should be considered. These comparative analyses also should

discuss such factors as changes in neutron flux densities or shifts in neutron flux

Spcytra

Fission-product inventory, barriers, and retention for the fuel should be analyzed.

These characteristics may differ significantly between the current HEU and the

proposed LEU fuel designs. For example, the licensee should demonstrate that the

fuel design will acceptably accommodate internal pressures from fission product

gases associated with the burnup of LEU fuel.

Safety limits and their bases for the HEU and LEU fuels should be compared, and

the LEU values should be shown to be applicable to the fuel design. These

parameters should be correlated to the dynamic design and the accident analyses.

Analyses for this section may simply invoilve reference to the appropriate section of

the SAR and report (i.e., NUREG-1281, -1282, and -1313) for the accepted fuel

design

4.2.2 Control Rods

Any changes, although not generally anticipated, to the characteristics of the

control rods (e g., the mechanical design, material, or configuration) for the

proposed LEU core should be described. An example of change in the control rod

design would be a change in neutron absorber material or distribution within the

control rod, or location in the core Such a change may be due to changes in the

neutron spectum for the LEU core. The changes should ensure that control rod

worth for the proposed LEU-fueled core remains within acceptable limits for

control or shutdown functions or for potential reactivity change accidents.

Another possible change would be the addition of fuel follower control rods to

maintain nuclear capability

4.2.3 Neutron Reflector

Any changes in the neutron reflector required by the conversion, including

mechanical and functional designs, materials of construction, thermal capabilities,

design life, and replacement procedures, should be discussed and analyzed in this

section of the SAR. The reflector design may need to be changed to maintain

thermal neutron flux at experimental facilities at HEU levels for the LEU cores.

NUREG- 1537, PART I 6 REv 0,2/96 NUREG- 1537, PART I 6 REV 0, 2/96
.lEU TO LEU CONVERSIONS

4.2.4 Neutron Source and Holder

Any differences in the neutron source design or location between the BEU and the

proposed LEU core should be analyzed to ensure continued safe function.

Although not generally expected, changes in fue geometry for the LEU core may

make it necessary to change the source, source holder, or source location.

4.2.5 h-Core Experimental Facilities

Any proposed changes to the in-core expeimental facilities required by converion

to the LEU fuel should be described. The design bases, design parameters, and

effects on reactivity during allowed operations should be compared for the HEUand LEU-fueled cores. This comparison should include the thermal-hydraulic

characteristics of the in-core experiment facilities and analysis of the therialhydraulic effects on nearby fuel elements. Accident scenarios related to in-core

experimental facilities should be presented in Section 13 on accident analyses. Ihe

use of the experimental facilities should be addressed in the experimental prdgrams

portion'of the HEU-to-LEU fuel-conversion SARL'

4.2.6 ReactorMaterials -

In this section of the SAR, the licensee should give the composition'of the LEU

fuel (e g., cladding and matrix material). Any chaniges in maftesial and

environmental conditions that are needed for the conversion from HEU to LEU

fuel also should be discussed. Such changes would include i~diation intensity''

changes as a result of neutron spectrum haidening, erosio*, corrosion,'and changes

as a result of fuel element diziension and associated flow rate changes. -

4.3 Reactor Tank and Biological Shield'

Any proposed modifications to thei tank/pool or the biological shield required by

the conversion from HEU to LEU fuel should be analyzed in the SAR. Although

not generally expected, these'components could change because of modifed'

geometry or neutron flux for the LEU-fuiled core. ' The licensee should -show that

the'charige in neutron spectrum should not cause unacceptable irradiation effects

on the tank/pool or the biological shield The licensee should demcnstrate that

modifications to the biological shield should not significantly decrease its physical

integrity against cracking or other stnictural mialfunc'tions for design 'conditions,

including natiral phenomena and accident co'nditions. The license'e'should show

that the radiation shielding' protection provided for the proposed LEU core and

reactor should not be significantly decreased below currently acceptableV'alues

REv 0,2/96 7 STANDARD FORMAT AND CONTENT

REV 0. 2/g6 7 STANDARD FoRmAT AND ComEN-r
APPENDI 18 1

4.4 Core Support Structure

The design features, materials of construction, criteria for rigidity and weight

support capability to maintain fuel, and other core components in position as

designed should be analyzed. The LEU core support structure should be

compared with that of the HEU core support structure in such areas as design

capability and material composition The licensee should verify that the design of

the core support structure remains valid for the LEU-fueled core or should

propose changes so that the design function is acceptably maintained. Specifically,

the changes in LEU fuel configuration, weight, and neutron spectrum or flux level

should be considered to ensure that the current core support structure or proposed

changes would perform the intended function.

4.5 Dynamic Design

The licensee should confirm that the integrity of the proposed LW fuel will not be

compromised The control systems and reactor physics parameters should be

analyzed and compared with the existing HEU-fueled reactor. Computed HEU

characteristics should be compared with BEU core measurements to assist in

confirming the validity of calculational methods and codes, and LEU

characteristics should be predicted using the same methods and codes These

analyses and comparisons should demonstrate that the reactor design, including

temperature and power coefficients, will prevent fuel failure from any allowable

power, coolant flow, and addition of reactivity conditions (accident conditions

should be assessed in the accident analyses, Section 13). The pulsing

characteristics of the reactor also should be analyzed if the reactor is authorized to

operate in that mode. Additional guidance may be found in selected portions of

NUREG-1281, -1282, and -1313, Bullock, 1962, Oak Ridge National Laboratory

reports ORNL- 2892 and TM-627; and Shibata, 1984.

4.5.1 Control Rod Worth and Excess Reactivity

The HEU-fueled core should be compared with the proposed LEU-fueled core for

control rod worth and excess reactivity. This comparison should include such

factors as the effects of xenon-135, samarium-149, and other fission products; void

coefficient, temperature coefficients of reactivity for fuel and moderator, burmup

and creation of fissile material; movable and secured experiments; and burnable

poisons. The neutron spectral hardening associated with conversion to LEU

should also be considered. Any changes in control rod worth or excess reactivity

should be analyzed, and required changes to associated technical specification

limitations should be established and justified Changes should be analyzed to

demonstrate that reactor control and function should be acceptably ensured The

control rod worth and excess reactivity should be verified in the startup test

NUREG-1537, PART I 8 REv 0.2/96

NUREG- I 537, PART I 8 REV 0, 2/96
.. , . . . HEU ToLEU CoNvWsIos

program .by'measurement. A range of expected values winth'acceptance criteia and

bases should be provided in this section and referenced to the startup test program.

4.5.2 Shutdown Margin

Any changes in the shutdown reactivity for the cold, clean reference core or any

other anithorized core arrangements and operating conditions'that may be limiting

with regard to shutdown margin should be analyzed.: The safe shutdown of the

reactor should be demonstrated, including any reactivity changes caused by

expected burnup and by failure of any allowed movable experiment. The

shutdown margin required by the technical specifications for the HEU-fueled core'

should be verified to be applicable for the LEU-fueled core or a change in the'

associated technical specifications and bases'should be proposed and justified.

Any change should be demonstrated to continue to ensure acceptable reactor

shutdown characteristics, for example, those specified in Chapter 14 as revised

from ANSI/ANS 15 1.

The shutdown reactivity should be verified to satisfy shutdown margin

requirements during the startup test program. A range of expected values with

acceptance criteria and bases should be provided in this section and'referenced to

the startup test program

4.5.3 Other Core Physics Parameters

The anticipated inherent reactivity feedback coefficierits, including the reactivity

coefficients of the fuel temperature, moderator temperature, moderator density and

voids, and the power defect should be provided. The HEU-fueled 'core physics

parameters should be compared with the LEU-fueled cores Thelcharacteristics

and mechanisms of a prompt fuel temperature coefficient of reactivity and its effect'

on stability'and safety of reactor operation, including any applicable'pulsing,

should be considered Changes in'delayed neutron fraction and prompt neutron

lifetime'resulting from the core conversionr should also be considered. 'These '

analyses should show that design parameters will compensate for any allowable'.

addition'of excess reactivity so that reactor control will be ensured. For example,---

a rapid reactivity change equivalent to the maximum allowable worth movable . '

experiment will not result in an uncontrollable reactor configuration

Thei'plutonium pr6duced during burnup of the proposed LEU fuel should be.

anAlyzed. The significant effects of plutonium-239'on LEU reactor operating

characteristics, such'as changes in reactivity and delayed neutron fraction, should

be included in these'analyses. The plutonium production should be compared with

existing HEU fuel, and methods to assess plutonium inventory, including end-oflife conditions, should be established and discussed The changes in plutonium

inventory for the LEU-fueled core should be considered and a brief description of

R ,29 9. SNDR .,OMA A_ , _ .E_

REV 0, 2/96 9 STANDARD FORMAT AND CONTEN'T
- - AAPPENDIX 18.1

the methods to ensure compliance with the requirements of 10 CFR 70 51 and

74.13 for material reporting and status reports should be provided (increased

reporting may be required for the LEU fuel under these regulations).

This discussion should compare the HEU and LEU core physics parameters in

tabular form. Any related license requirements or proposed changes in technical

specifications required by conversion from HEU to LEU fuel should be discussed

and justified, and reference should be made to the discussion on license or

technical specifications changes.

Core physics parameters that verim* proper core configuration and behavior and

can be accurately measured, should be verified in the startup test program. A

range of expected values with acceptance criteria and bases should be provided in

this section and referenced to the startup test program for comparison with startup

test results.

4.5.4 Operating Conditions

The dynamic operating conditions for the LEU fuel designs should be compared to

those for the HEU fuel. Core thermal power and neutron flux spatial distributions,

both radial and axial, power peakdng within individual fuel plates or rods, and

related temperature distributions for the range of allowable operating conditions

should be analyzed. This analysis should include calculations of both the BEU and

LEU cores, giving a quantitative description of the main changes in operating

characteristics. The values and characteristics calculated as operating conditions

(e.g., maximum temperatures, power peaking factors) should be used or

referenced in other areas such as under accident analyses in Section 13.

The dynamic response of the reactor to anticipated and postulated disturbances in

the process variables and to changes in reactivity, including inherent feedback

mechanisms and protective actions of the reactivity control elements, should be

provided. For example, analyses for maximum allowable power, temperature, and

reactivity addition conditions should not result in exceeding LEU fuel design

temperature limits Further, any relevant changes in the technical specifications

should be discussed and justified here and referenced to the technical specifications

discussion.

Operating parameters that verify proper core configuration and behavior and that

can be accurately measured should be verified by measurements in the startup test

program. A range of expected values with acceptance criteria and bases should be

provided in this discussion and referenced to the discussion of the startup test

program.

NUREG- 1537. PT I 10 REv. 0,2196

NUREG- 1537, PART I 10 RE%'. 0. 2/96
HEU TO LEU CoNvERSIoNS

4.6 Functional Design of the Reactivity Control System

Although not expected for LEU conversion, any proposed changes in the design

features of the rod-drive systems should be analyzed. Potential changes include

reactivity worth and rate of addition or withdrawal for allowable conditions of the

LEU-fueled cores. The design of the reactivity control system should be inalyzed

to verify function for the LEU core under postulated scenarios and conditions of

the accident analyses.

Selected parameters, such as those described in Section 12.6, 'Reactor Reload and

Startup Plan," should be established to verify proper core configuration and..

behavior through measurement in the startup test program -A range of expected

values with acceptance criteria and bases should be provided in this section and

referenced to the startup test programThe analyses of any necessaiy changes in the technical specifications that address

reactivity control systems should be provided and justified in this section and

referenced in the specific discussion on technical specifications changes.

Additional guidance is provided in selected portions of ANSI/ANS 15.15-1978

and ANSI/ANS 15.20 (draft)

4.7 Thermal-Hydraulic Characteristics

Any changes in the thermal-hydraulic parameters and characteristics during routine

and transient conditions that may anse from the fuel conversion should be

analyzed. Possible changes include the number or dimensions of coolant channels

,and fuel plates or rods, core dimensions, pow' density, fuel and'cadding

temperatures, surface heat flux, radiation heating, thermal conductivity of the fuel,

and coolant flow rate. Changes to powerpeaking factors for the conversion to the

LEU reactor core should be included for expected and linu-ting control rod

operating conditions in this analyses This'analysis should show that the LEUfueled reactor thernal-hydraulic design will finction under postulated accident -

scenarios and conditions.

The HEU and LEU thermal-hydraulic characteristics should be calculated. The

analytical methods may be confirmed with measurements on the licensed HEU

core Any change in the therial-hydraulic parameters between the HEU core and

the proposed LEU'core should be discussed. Changes in the therial-hydraulic

characteristics when converting from HEU to LEU fuel should not result in

exceeding design limits, such as departure from nucleate boiling ratio (DNBR),

flow instability,-'or fuel safety limits. If there would be significant changes in fuel

element' flow rates or heat transfer characteristics, verification should be provided

that the changes result from factors in the conversion process beyond the control

REv. 0,2/96 11 STANDARD FoIu.vT AND CO'TENT

REn'. O. 2/96 I I STANDARD FoRmAT AND Co%-rENT
APPENDI 18.1

of the licensee. Any proposed changes to the technical specifications, including

safety limits, should ensure that the reactor will operate safely during routine and )

transient conditions under all analyzed combinations of system parameters.

The initial startup test program should include comparisons, confirmations, and

acceptance critenia for thermal-hydraulic characteristics that verifr proper core

configuration and behavior and that can be accurately measured& A range of

expected values with acceptance criteria and bases should be provided in this

section and referenced to the startup test program.

5 REACTOR COOLANT SYSTEMS

Any changes in the reactor coolant system or associated procedures that are

required by the conversion to the proposed LEU fuel should be analyzed. This

analysis should contain a summary functional description of the individual

components and a statement of relevant characteristics (e.g., pressure,

temperature, and flow rate). The mechanisms and system characteristics that

provide for the automatic transition of forced-flow to naural-flow convection

cooling as required to maintain fuel integrity should be analyzed. Changes in the

reactor coolant and associated systems needed to convert from HEU to LEU fuel

should not result in exceeding any design or safety limit. Changes in this system

may affect the reactor, engineered safety features, instrumentation and controls,

and accident analyses. For example, changing fuel geometry may result in changes K>

in reactor coolant system flow rates and pressure drops, which may require

changes to reactor coolant system configuration or components.

The bases and justification for any changes that may be needed in the related

technical specifications should be analyzed in this section and ieferenced to Section

14 on technical specifications. Selected parameters that verify proper core

configuration and behavior and that can be accurately measured should be verified

by measurement in the startup test program. A range of expected values with

acceptance criteria and bases should be provided in this section and referenced to

the startup test program.

6 ENGINEERED SAFETY FEATURES

Any changes in the engineered safety features (ESFs) required to accommodate

new or changed accident progressions brought about by the proposed conversion

from HEU fuel should be analyzed. This analysis should consider any accident

analyses assumptions and conditions for the ESFs. ESFs may include the reactor

room ventilation system, the stack exhaust system, and an emergency core cooling

system (ECCS). Any modifications or additions of ESF systems or functions

should ensure that design and safety limits continue to be satisfied. Such changes

may use guidance from Chapter 6 of the format and content guide.

NUREG-1537, PART 1 12 REv 0.2/96

NUREG- 1537, PART I 12 REV 0, 2/96
-.- - EU To LEU CovwvEsoNs

The bases for any changes that may be needed in the related technical'

specifications should be analyzed and discussed or referenced to that topic.

Selected parameters that verify proper core configuration and behavior and that

can be accurately measured should be verified in the startup test program. A range

of expected values with'acceptance criteria and bases should bi'provided in this

section and referenced to the startup test program.

7 INSTRUMENTATION AND CONTROL SYSTEMS

Although changes are not expected for the conversion to LEU fiel, any changes in

instrmentation and control systems, including instrument readings, interlocks,.

inhibits, and trip setpoints, should be analyzed. .This analysis should demonstrate'

that there is no significant decrease in functional safety capabilities, including

speeds of response, redundancy and diversity, or reactor shutdown capability.

Additional guidance may be found in ANSI/ANS 15.15-1978 and ANSI/ANS

15.20 (draft).

8 ELECTRICAL POWER SYSTEMS

Although changes are not expected for the conversion to LEU fuel, any changes in

the electrical power systems required to implement the core conversion should be

analyzed This includes normal power systems and emergency or backup power.

Systems.

9 AUXILIARY SYSTEMS

Any changes in operational characteristics or in components of ainoiiary systems

required by the BEU to LEU conversion should be discussed in this'section. For

example, specific consideration should be given to fuel handling and storage

systems for conversion to LEU fuel Although noi expected, changes to the

ventilation and exhaust systems or the heating and air conditioning systems for the

reactor room and auxiliary rooms should be explained.

Any changes in the LEU fuel configuration or the radioactive source terms from

irradiated fuel elements and the effect on fuel storage, handling, and shipping.

should be analyzed The storage provisions for both fresh and spent fuel elements,

their locations, and their capacity should be presented. Criticality safety

considerations should be analyzed. For example, maximum k~ff under'optimum

moderator conditions should be determined to be within technical specifications,.

limits and SAR assumptions. If fuel movement and receipt are not regular..

occurrences, updated procedures may be required Any changes to the fuel

handling and storage systems or procedures should be shown to maintain safe

conditions, including shielding, subcritical configuration, and cooling.

REX' O. 2/96 13- STANDARD FoRMAT AND CoNTrENr
APEDIX 18. I -

If the LEU core. will be received before the disposal of the HEU core, storage of

both cores should be analyzed. The requirements and capability to maintain selfprotecting radiation levels of the HEU fuel should be considered.

Additional guidance may be found in ANSI/ANS 15.19-1991.

10 EXPERIMENTAL FACILITIES AND

UTILIZATIONAny changes in the expeimental irradiation filities required by the core

conversion, such as changes to thermal columns, neutron beam facilities or pool.

irradiations, as well as changes to iri-core experiment tubes and pneumatic transfer

systems, that were not analyzed elsewhere in the SAR, should be analyzed. If

changes were analyzed elsewhere in the SAR, this discussion should make

reference to those analyses. See the discussion in Chapter 10 of the format and

content guide for additional guidance on this subject

11 RADIATION PROTECTION AND RADIOACTIVE

WASTE MANAGEMENT PROGRAMS

Radiation protection and radioactive waste management programs overlap each

other on certain issues for non-power reactors. Guidance may also be found in

Chapter 11 of this document, Regulatory Guides (RGs) 1.25 and 1.109,

NUREG-0851, and ANSL/ANS 15.7-1977 and 15.11-1993. The licensee should

also provide its "as low as is reasonably achievable (ALARA) program and

methods for compliance with other applicable regulations, such as 10 CFR Part 71.

11.1 Radiation Protection Program

The following effects on the radiation protection program from modifications to

any components or systems, or from changes in operational limits or procedures

required by the conversion to LEU fuel should be analyzed:

* potential radiation sources and magnitudes

* radiation detection and measurement methods and procedures

* provisions made to assess potential exposures and doses to the staff and

public

* methods used to limit doses to within applicable regulations

NUREG-1537.PARTI 14 REV O2/96 , 14 REV 0, 2/96
HEU To LEU CoNv~psioNs

* commitment to an ALARA program for potential exposures

* .assessment of the effectiveness of the dose control programs

* methods to detect potential inidvertent criticality

11.2 Radioactive Waste Management

The following items should be addressed if affected by the conversion to LEU fuel:

* the capabilities of the reactor facility to control, collect, handle, process,

store, and dispose of liquid, gaseous, and solid wastes that nay contain'-

radioactive materials

* the instrumentation and methods used to measure the quantities of

radioactivity and to monitor the release of radioactive wastes outside of the

restricted area

All types of radioactive wastes that are generated (solid, liquid, and gaseous) in'the

operation and maintenance of the reactor facilities and in the experimental

programs should be considered. How the conversion of the fuel would change the'

formation, control, release, or disposition of radioactive waste materials'associated

with the reactor should be analyzed. Any changes'to the faclity or procedures

should be analyzed and their effects evaluated. Any changes in methods used to

store and dispose of radioactive wastes should be described. Any changes in the

methods used to assess the source strengths and predict potential radiation

exposures to personnel due to radioactive waste in both restricted and unrestricted

areas should also be described : - ,

For many non-power reactors, argon-41 is a significant radioactive waste that is

produced by operations. All changes in the sources of argon-4 and all potential

exposure pathways should be addressed. Changes in the sources of argon-41 may

be related to changes in fuel geometry and neutron spectra.

The handling or release of all radioactive wastes should be verified to remain

consistent with the facility program, applicable regulations, and the facility

guidelines with regard to ALARA.

12 CONDUCT OF OPERATIONS

This section provides information on planning and implementing the activities for.

reactor facility operation during conversion from HEU to LEU fuel.

. . . . _ : _5 REV 0 2/96 15, STANDARD FORMAT AND CO%'rENT
APPENDix 18.1

12.1 Organization and Staff Qualifications

It is not expected that reactor conversion from HEU fuel will require changes in

the licensee organization. Note that significant organization changes would have

already been reviewed and accepted by NRC in technical specifications and in the

license revisions.

12.2 Procedures

Any procedural changes necessary for the conversion from HEU fuel should be

described. Procedural changes to account for LEU-fueled core and fuel

configuration may include such issues as location of fuel elements and control

rods, conduct of critical experiments, storage and shipment of fuel elements as well

as procedural changes to account for needed technical specifications changes

The licensee should consider specific changes in shipping requirements and

procedures (e.g, type B material, and associated quality assurance program and

cask user registration requirements). ANSI/ANS 15.19-1991 provides guidance.

Any new procedures that are needed for the conversion should similarly be

addressed. For example, changes to the reactor coolant system to ensure adequate

flow may require modified operational procedures

12.3 Operator Training and Requalification

Certain operating characteristics, license conditions, operating procedures, and

technical specifications requirements may change as a result of thefuel conversion.

The implications of these changes should be incorporated into the operator training

and requalification program The operator training and requalification program

should clearly establish the procedures, techniques, and on-the-job training which

ensures that the licensed operators are instructed on all changes to the reactor, its

operating characteristics, written procedures, license conditions, and technical

specifications. Additional general guidance can be found in

ANSI/ANS 15 4-1988

12.4 Emergency Plan

The conversion to LEU fuel should not significantly affect the emergency action

level guidelines or potential reactor emergency event consequences. Although

changes are not expected, any changes to the emergency plan that might occur

should be submitted as part of the conversion application in accordance with

applicable regulatory requirements Additional general guidance may be found in

ANSIIANS 15 16-1982 and RG 2 6

NUREG-1537. PART I 16 REV 0, 2/96
lIEU To LEU CoNvE~siows

12.5 Physical Security Plan

A less demanding physical security plan could result from the conversion to LEU

fuel if the conversion affects the category for possession of special nuclear material

as discussed contain 10 CFR 73.2, 73.6, and 73.67. Any proposed changes in the

physical security plan and implementing procedures should demonstrate that the

applicable requirements of the regulations are still met when either irradiated or

unirdiated'fuel is at the facility and in transit. These changes should allow

possession of LEU and BEU fuel for as long as necessary. Mditionally, ifthe

licenisee has relied on the exemption of 10 CFR 73.6(b) for self-protection of the

HEU fuel, the licensee should demonstrate that the exemption will continue to be

applicable while the HEU is still on site or should discuss the options to the selfprotecting exemption. Applicable alternatives and changes to the security plan and

related, license conditions should be proposed insaccordance with regulatory

requirements.

12.6 Reactor Reload and Startup Plan.

Reloading with LEU fuel and implementing a startup plan are very similar to the

initial loading and startup as part of the original licensing action. Many of the

same concerns and precautions should be considered in the conversion reload. A

reload and startup plan should be submitted as part' of the proposal for

authorization to convert the reactor from the use of HEU fuel to make sure that

the operating characteristics are well understood and to validate the predicted

behavior. Therefore, if practical, measurements of selected parameters' of the

HEU-fueled core should be compared to calculated values to verify inalyticii

methods and ensure that meaningful acceptance criteria for the LEU-fueled core

have been established from the calculational methods' The plf -should provide

that the measurements for the LEU reactor are compared to acceptance criteria as

established in this conversion proposal. The acceptance'criteria should ensure that

the LEU reactor is functioning as it was designed and analyzed and that the license

and technical specifications will be satisfied

The plan should contain the following items:

* a well-planned systematic set of subcritical multiplication measurements or

an inverse multiplication approach to critical measurement during new fuel

loading, and confirmation that analysis subcritical multiplication or critical

fuel loading are within preestablished acceptable limits

* an experimental measurement plan to determine the important operational

reactor physics parameters (such as control rod worth, excess reactivity,

reactor thermal power, coefficients of reactivity, and power peaking factors)

REv 0.2/96 17 STANDARD FORMAT AND CON7ENT

REV' O. 2196 17 ; STANDARD FoRmAT AND Co.N7ENT
AwEmix 1 8.1

and thermal-hydraulic parameters (such as fuel, cladding, and coolant

temperatures, reactor coolant system flow rates, and pressure drops, if

appropriate), comparisons with predictions and acceptance criteria

established in the applicable section of this conversion proposal, and

discussions of any discrepancies that may have arisen

*measurenents of magnitudes of area radiation fields and radioactive,

effluents, and comparisons with the same parameters for operation of the

lIEU-fueled reactor and preestablished acceptance criteria from the

applicable sections of this conversion proposal for the LEU-fueled reactor

A startup report should be submitted to the NRC, in accordance with the time

frame required in the technical specifications or within 6 months after the LEU fuel

is loaded. An example of an outline for a startup report is given below:

(1) critical mass and final criticality conditions for this first LEU core

* measurement with HIEU cores

* measurement with LEU cores

* comparisons with acceptance criteria established earlier in

conversion proposal calculations and analyses

(2) measurements and comparison to prediction of subcritical multiplication for

initial core fuel loading

(3) excess (operational) reactivity

* measurement with HEU cores

* measurement with LEU cores

* comparisons with acceptance criteria established earlier in

conversion proposal calculations and analyses

(4) control and regulating rod calibration

* measurement of differential and integral rod worth for HEU core

* measurement of differential and integral rod worth for LEU core

* comparisons with acceptance criteria established earlier in

conversion proposal calculations and analyses

NUREG- 1537, PARJT I18RV029 18 REV 0 2/96
lHEU TO LEU CoNvEiMSoNs

(5) reactor power calibration

* methods and measurements that ensure operation within the license

limit

* comparson between BEU and LEU nuclear instrumentation

setpoints, detector positions, and detector output

comparisons with acceptance criteria established earlier in

conversion proposal calculations and analyses

(6) shutdown margin

* measurement with BEU

measurement with LEU

* comparisons with acceptance criteria established earlier in

conversion proposal calculations and analyses

(7) partial fuel element worth

* measurements of the worth for lEU-fueled core

* measurements of the worth'for LEU-ffieled core

* comparisons with acceptance criteria established earlier in

conversion proposal calculations and analyses

(8) thermal neutron flux distributions

* measurements of core and experinental facilities with HEU

* measurements of core and e'cpimental facilities with LEU

* comparisons with acceptance criteria established earliei in'

conversion proposal calculations and analyses

(9) enhanced radiation measurements of reactor coolant fission product

inventory or release during startup test program to detect potential fuel

fission product barrier degradation or contamnination from'other' sources

* - measurements of core and experimental facilities with HEU

REV 0, 2-96 19 STAEDARD FORMAT AND CONTENrT -
atAPPEND:i 18 1

* measurements of core and experimental facilities with LEU

* comparisons with acceptance criteria established earlier in

conversion proposal calculations and analyses

(10) results of determination of LEU effective delayed neutron fraction,

temperature coefficients, and void coefficients for LEU- and HEU-fueled

cores and comparison with acceptance criteria established earlier in

conversion proposal calculations and analyses

(11) comparison of the various results with the two fuel types, and

discussion of the comparison, including an explanation of any

significant differences that could affect operation and possible

accidents with the reactor

(12) thernal-hydraulic characteristics such as fuel, cladding, and coolant

temperatures, and reactor coolant system flow rates and pressure drops

13 ACCIDENT ANALYSES

In this section, the licensee should demonstrate that the accident analyses results

for the LEU reactor do not significantly exceed acceptance criteria of the accidents

previously postulated and analyzed in the safety analyses or evaluated in staff

safety evaluation reports In the accident analyses, the licensee also should

consider if any new accidents may be introduced by the conversion.

This discussion should summarize the comparison of the HEU and LEU accident

analyses. This comparison should demonstrate that the conclusions previously

reached in the safety analyses and staff safety evaluation reports remain valid for

the LEU-fueled core. For example, a maximum hypothetical accident that assumes

a fission product release scenario for an HEU fuel element may be used for the

LEU fuel if the licensee demonstrates that the two fuels have sufficiently similar

physical characteristics and fission product inventories. If these conditions can be

verified, a largely qualitative analysis with check calculations should be sufficient.

This process involves a comparison of the scenarios and conclusions for the HEUfueled reactor accidents considering the physical characteristics of the LEU-fueled

reactor

New or revised accident analyses for the LEU fuel should be performed if the

techniques used for the current HEU analyses are no longer available or

appropriate, or if the conversion creates the possibility of a new, previously

unanalyzed accident scenario or significantly changes the potential consequences

of a previously accepted HEU accident scenario (If this approach is adopted, the

licensee may need to reanalyze some HEU data with the new methods to provide a

_UE-57 _AR _ 20 RE .0A2A9

NUREG- 1 537. PART I 20 REV 0, 2/96
HEU TO LEU CONVERSIONS

valid comparison) 'New vor revised safety analyses should demonstrate that the

LEU core can safely withstand the accident. That is, these aialyses should show

that the consequences meet the acceptance criteria that were used and accepted for

the HEU-fueled reactor. Alternatively, these analyses should show that the: -.

accidents meet the acceptance criteria that are applicable to current non-power

reactor licensing activities (e.g., 10 CFR Parts 20 or 100). ,Ifnew safety analyses

are required for the conversion, all enveloping scenarios should be identified and

analyzed for the LEU reactor.

13.1 Maximum Hypothetical Accident (MHA)

The MHA is the hypothetical accident in which the potential radiological

consequences to the public health and safety are greater than those from any other.

postulated event at the facility The MHA has the greatest radiological

consequences of analyzed accidents for a non-power reactor to the facility staf

the public, and the environment The MHA is usually, but not always, associated

with assumed fuel cladding or fission product retention failure and fission product

release. This accident usually assumes conditions that are not considered credible,

but are bounding and demonstrate that under the most extreme conditions and

assumptions, the radiological consequences at a non-power reactor could not

exceed previously used acceptance criteria (e.g., 10 CFR Part 20 or 100):.

If the HEU and the LEU reactor characteristics are sufficiently similar, principally

in physics, thermal-hydraulics, and radioactive material inventories, the licensee

may be able to demonstrate that the conversion from HEU to LEU fuel does not

introduce a new and unreviewed type of MHA and does not invalidate the,.

conclusions about HEU reached in the SAR and in the staffs safty. evaluations on

radiological consequences of the MHA As previously discusod for MHA

scenarios, the licensee may use comparative analyses with relatively simple

calculational checks to demonstrate that conversion causes no significant increase

in the radiological consequences of the MHA . -

On the other hand, if the conversion does create the possibility of a previously

unanalyzed MHA scenario or significantly increases the potential consequences of

the previously accepted MHA, safety analyses should be submitted for the:

accident., In this case, the MHA analysis should demonstrate adequate assurance

of public health and safety for the proposed LEU core and fuel.- That is, the MHA

analysis should show that the appropriate acceptance criteria for non-power

reactor accident analyses continue to be met or that additional requirements or

compensatory measures are proposed to ensure that the acceptance criteria will be

met. .

The probability or mechanistic details of the MHA scenario need not be described

or evaluated Only the consequences need be described and analyzed. MHAs are

REV.O,2/96 21 STtn FORMAT AND CoN'1mr

RE%,. 0. 2/96 21 STANDARD FORMAT AND CoN-mNu
aAPE~NDIX8 1

facility specific. Additional general guidance may be found in RGs 1.25 and 1.109,

NUREG&0851 and -1313, ANSI/ANS 15.7-1977, an NRC memorandum dated . _)

June 17, 1981, U.S. Atomic Energy Commission report TID-14844, and in

Chapter 13 of this format and content guide.

13.2 Rapid Addition of Reactivity Accident

This accident analysis assumes the addition of the maximum reactivity for a single

credible event to the reactor. It also assumes that this maximum reactivity is added

to the reactor at the maximum credible rate. Any feedback effects (e.g., the

Doppler effect in uranium-238) changed by converting to LEU should be

considered. As for other accident analyses, this accident analysis may be carried

out by one of the following two methods:

(1) The parameters that are important to this accident should be compared to

verify that the LEU fuel and core will not differ significantly from the same

accident analyzed for HEU fuel and core. The limiting rapid addition of

reactivity accident scenario postulated in the current SAR for HEU fuel

should be described, as well as the initial core and reactor conditions that

are pertinent to the analysis methods and assumptions for this accident

Using the appropriate analytical methods, the licensee should verify that the

consequences arising from this accident scenario for the LEU-fueled core

will not rwesut in a significant change from that for the HEU fuel (i.e., there

will be no significant reduction in the margin of safety to maintaining fuel

integrity) Methods of analyses for licensee consideration are discussed in

an article by William L. Woodruffthat appeared in February 1984 in the

journal Nuclear Technology

(2) If a meaningfil comparison is not feasible because the calcilacional

techniques have changed since the current SAR analyses for HEU fuel

were performed or because the LEU fuel and core characteristics cannot

be correlated and shown to be similar to that of the HEU, a new

standalone analysis for LEU fuel assessing the rapid addition of reactivity

accident safety should be done This new analysis should describe the

limiting accident scenario, stating the postulated initial core and reactor

conditions that are pertinent, and the analysis methods and assumptions to

be used. The analysis of this accident scenario for the LEU-fueled core

should demonstrate that the effects of this accident will not result in any

localized fuel melting or any challenge to the cladding integrity or fuel

fission product retention capability, or, alternatively, will not increase the

potential radiation exposure over that acceptable for non-power reactors

For validation of the method, analysis of HEU fuel and core conditions

should be considered, if applicable, for comparison to the previously

NUREG-1537,PART I 22 REV 0,2/96 NUREG-1537, PART I 22 REV 0, 2196
HEU TO LEU CONVERSIONS

reviewed and accepted analysis of the accident involving a rapid addition of

reactivity

133 Reduction-in-Cooling Accidents

Two such accidents are the loss-of-coolant accident (LOCA) and the loss-of-flow

accident (LOFA). The scenario(s) as postulated in the current SAR and any

diffeeces in the progression of the accidents with LEU fuel as compared to HEU

fuel should be described. Any mitigating effects provided by an ESF,-such as the

ECCS, should be analyzed.

Comparison of the parameters of importance to these HEU and LEU accident

analyses should verfy that the temperature changes over time in the applicable

LOCA and LOFA scenarios do not result in safety analysis findings greater than

the acceptance criteria for HEU and LEU fuel and cores. The postulated initial

core and reactor conditions that are pertinent to the comparson HEU of the and

LEU cores and the thermal-hydraulic methods and assumptions should be'

described. This analysis should be conducted to the point at which LEU fuel

cooling and shielding conditions have been stabilized The analysis should

demonstrate that the consequences to fuel integrity and the intensity of the

radiation field for the LEU-fueled core are not significantly increased from those

acceptable for the licensed lEU-fueled core.

I, however, the conversion significantly increases the consequences of the LOCA

or LOFA for the LEU-fueled core or introduces a new and previously unanalyzed

accident scenario, new safety analyses should be submitted. The consequences

arising from these scenarios that involve reduced cooling should not result in any

localized fuel melting or breach of cladding integrity or fuel fission product

retention capability, or, alternatively, should not significantly increase the potential

radiation exposure over the exposure that is acceptable for the non-power reactor

accident analyses. Additional guidance may be found in Hunt and DeBevee, 1959;

Knexevich et al., 1965;'Guilf Gneral Atomics GA 6596, and Chapter 13 of this

document.

13.4 Other Accidents

The accident caused by the failure of the fuel fission product barrier generally

involves the failure of fuel plate or rod cladding under some mechanistic scenario.

As previously discussed, thisiaccident scenario may be the MiHA for some nonpower reactors It may also assume failures of large fractions of the fuel fission -

product barrier in 'order to demonstrate acceptable safety design for a broad variety

of potential events and conditions. -,The potential radiological consequences to the

facility staff, the public,' and the environment should be evaluated using

conservative dose calculation methods and assumptions

R ~.0. ./9 23 . .. S.NDR ..... FO.A AN 

REX'. 0, 2/96 23 STANDARD FORMAT AND CONSENT
APPENOD: 18.1

As discussed in Section 13.1 of this appendix, if the HEU and the LEUJ reactor

characteristics are suiciently similar, the licensee may be able to demonstrate that

the conversion does not introduce a new and unreviewed type of such an accident

and does not invalidate the BEU SAR and staff conclusions on the associated

radiological consequences. The licensee may use comparative analyses with

relatively simple calculational checks to demonstrate that conversion causes no

significant increase in the radiological consequences of the fission product barrier

failure. On the other hand, if the conversion does create the possibility of a

previously unanalyzed scenario or significantly increases the potential

consequences of the previously accepted accident analyses, a revised SAR should

be submitted for the accident to provide reasonable assurance that the radiological

consequences will not endanger the facility staf the public, and the environment

Other postulated accidents may be affected by or result from the conversion from

HEU to LEU fuel. Such reactor-specific accidents as the following should be

identified and analyzed:

e misplaced experiments or experiment malfunctions (considering different

reactivity response for the LEU and HEU fuel)

* mishandling or malfimction of fuel (considering potential differences in

radioactivity release for LEU and BEU fuel) <2

* loss of normal electrical power

* external events

* mnishandling or malfunction of specific equipment

* leakage of coolant into the cladding with a resulting steam pressure failure

of the cladding

* failure of a fueled or unfuieled experiment

* physical damage to the core

14 TECHNICAL SPECIFICATIONS

Changes in the tichnical specifications that are required by the conversion should

be proposed and explained. Technical specifications that may need to be changed

are those specifications and their bases that relate to the fuel (e.g., core reactor

physics and thermal-dynarnic parameters dependent on fuel chiracteristics, and fuel

storage requirements), including safety limits and limiting safety system settings. * }

NTJREG.1537. PART) 24 REv 0,2/96 NUREG- 1 537, PART I 24 REV 0, 2/96
lIEU To LEU Co1~vERsio~js

In addition, the technical specifications design features that describe the core and

fuel parameters may need to be modified.

The changes in the technical specifications should be the minimum required by the

differences in the characteristics of the HEU and LEU fuels. These changes should

be traceable to the BEU to LEU conversion safety analyses from which they are

derived. Any new or modified structures, systems, or components required for the

BEU to LEU conversion should be added or appropriately changed in the

applicable'sections of the technical specifications (e.g., new or modified ESFs).

The licensee should supply appropriate technical specifications consistent with the

assumptions of the LEU safety assessment. Technical specifications should.

contain the necessary requirements specified in 10 CFR 50.36. The format and

content of the revised technical specifications should be maintained the same as the

current technical specifications. If a change in content of the technical

specifications is appropriate for.the LEU conversion, additional guidance is

provided in ANSI/ANS 15.1-1990 and Chapier 14 of this document..

15 OTHER LICENSE CONSIDERATIONS

15.1 Prior Utilization of Reactor Components

If the BEU fuel will be replaced by new, unadiated, ricently fabricated LEU fuel

that has been approved by the NRC (such as uanium silicide-aluminum dispersion

fuel or high-uranium-density, 20-percent-enriched TRIGA fuel), this discussion

may not be applicable.

If the replacement fuel is not new (e g., ifit has been irradiated, suchias previously

exposed TRIGA fuel), or if it was not recently fabricated (such as'SPERT fuel),

histories and safety implications of using the LEU fuel should be analyzed. This

analysis should consider custodianship; earlier operations; environmental

conditions under which the fuel wasused and stored,'operatiouial factors such as

power level, neutron flux and fluence level, burnup, temperatures and temperature

gradients across the fuel and the cladding, and coolant flow rates; pulsing; and

chemical composition and temperature ofthe coolant in which the fuel resided

over its lifetime

The effect of these conditions on fuel integrity should be analyzed, including

consideration of irradiation and test experience for similar or identical fuel; erosion.

and corrosion resulting from previous use; internal pressure and composition of

gas, including fission product gases; fuel inspection results and plans; and startup

test plans and acceptance criteria established earlier in this conversion proposal.

RE 0_/62 TNADFRA N orw REX'. O. 2/96- 25 'STANDARD FoRmAT AND CoNTENT
--

APIENDDC 18 1

If appropriate,. prior use of other components required for the conversion to LEU

(e.g., experimental facilities that are compatible with fuel design) or of existing

reactor components should also be analyzed.

15.2 License Conditions

The need for revised license conditions for the possession of the LEU fuel should

be considered since the LEU fuel and core will in all likelihood have a different

amount of uranium-235 than the HEU fuel and core. To arrive at a 'reasonable

possession' limit, the amount of LEU fuel assemblies that are needed to operate

the reactor plus an adequate nimber of LEU fuel assemblies to accommodate the

time necessary to get additional fuel to compensate for burnup, fuel failure, or

other reasons to replace fuel should be explained. An example of a license

condition for possession is the following:

pursuant to the Act and 10 CFR Part 70, 'Domestic Licensing of

Special Nuclear Material to receive, possess and use up to

kilograms of contained uranium-235 at enrichments less than

20 percent in the form of non-power reactor fuel in connection with

operation of the reactor, and to possess, but not separate, such special

nuclear material as may be produced by the operation of the facility.

The licensee should propose similar wording or the wording in the current license

for possession of material. It should be noted that other special nuclear material

(e.g., sources) must be specified in the license to allow possession.

A license change to possess but not use the HEU fuel until it can be shipped off

site and is no longer under the jurisdiction of the reactor licensee also should be

proposed if the plans are to possess both HEU and LEU fue sinultaneously. An

example follows:

...pursuant to the Act and 10 CFR Part 70, 'Domestic Licensing of

Special Nuclear Material," to possess, but not use, up to

xx kllograrms of contained uranium-235 at equal to or greater than

20 percent enrichment and other such special nuclear material produced by

operation of the facility in the form of non-power reactor fuel until this fuel

is removed from the facility.

15.3 Decommissioning

The effects of conversion to LEU should be evaluated. Changes may involve

activation changes due to spectrum hardening and flux distribution changes.

Additional guidance in this regard may be found in Chapter 17 of the format and

content guide

NtJREG.1537.P4'iT I 26 REV 0.2196

NUREG- 1 537. PART I 26 REV 0, 2196
- . HEU TO LEU CoNvERsioS

15.4 Bibliography

Aicerican National Standards Institute/American Nuclear Society,

ANSItANS 15.1, The Development of Technical Specifications for Research

Reactors," ANS, LaGrange Park, Illinois, 1990.

American National Standards InsfitutelAmerican Nuclear Society, ANSVANS

15.2, "Quality Control for.Plate-Type Uranium-Aluminum Fuel Elemens,"

ANS, LaGrange Park, Illinois, 1990.

American National Standards nstitute/American Nuclear Society, ANSVANS

15.4, "Selection and Training of Personnel for Research Reactors," ANS,

LaGrange Park, Illinois, 1988.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.7-1977, "Research Reactor Site Evaluation," ANS, LaGrange Park, Illinois,

1977.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.11, "Radiological Protection at Research Reactor Fcilities," ANS, LaGrange

Park, Illinois, 1987.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.15, "Criteria for the Reactor Safety Systems for Research Reactors," ANS,

LaGrange Park, Illinois, 1978.

American National Standards Institute/American Nuclear SQciety, ANSI/ANS

15.16, "Emergency Planning for Research Reactors," ANS, LaGrange Park,

Illinois, 1982.

American National Standards Institute/American Nuclear Society, ANSV/ANS

15.19, "Shipment and Receipt of Special Nuclear Material (SNM by Research

Reactor Facilities," ANS, LaGrange Park, Illinois, 1991.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.20 (draf) "Criteria for the Control and Safety Systems of Research Reactors,"

ANS, LaGrange Park, Illinois.

Bullock, J. B., "Calculation of Maxdmum Fuel Cladding Temperatures for Two

Megawatt Operation of the Ford Nuclear Reactor,". Memorandum Report No. 1,

Memorial Phoenix Project, Michigan, June 1962.

Gulf General Atomics, GA 6596, J. R. Shoptaugh, Jr., "Simulated Loss-of-Coolant

Accident for TRIGA Reactors," August 18, 1965.

REv 0,2/96 27 STANDARDFORMAT D CONTENT
AFVENDix 18.1

Hunt, C. H., and C. J. DeBevee, GEAP 3277, "Effects of Pool Reactor

Accidents," General Electric Technical Information Series, November 2, 1959.

International Atomic Energy Agency, LAEA-TECDOC-223, "Research Reactor

Core Conversion From the Use of Highly Enriched Uranium to the Use of Low

Enriched Uranium Fuels-Guidebook," Vienna, Austria, 1980.

International Atomic Energy Agency, IAEA-TECDOC-324, 'Research Reactor

Core Conversion From the Use of Highly Enriched Uranium to the Use of Low

Enriched Uranium Fuels-Guidebook Addendum on Heavy-Water-Moderated

Reactors," Vienna, Austria, 1985.

International Atomic Energy Agency, IAEA-TECDOC-643, "Research Reactor

Core Conversion Guidebook," Vienna, Austria, 1992.

Knexevich, M., et al., 'Loss of Water at the Livermore Pool Type Reactor,"

Health Physics IL pp. 481-487, 1965.

Nyer, W. E., et al., "Transient Experiments With SPERT-1 Reactor," Nucleonics,

June 1956.

Oak Ridge National Laboratory, ORNL 2892, J. F. Wett, Jr., "Surflace

Temperatures of Irradiated ORR Fuel Elements Cooled in Stagnant Air,"

April 16, 1960.

Oak Ridge National Laboratory, ORNL TM-627, T. hi Sims and W. IL Tabor,

'Report on Fuel-Plate Melting at the Oak Ridge Research Reactor, July 1, 1963,"

October 1964.

Shibata, Toshikam, et al., "Release of Fission Products From Irradiated Aluminide

Fuel at High Temperatures," Nuclear Science and Engineenng 87,. pp. 405-417,

1984.

Thompson, T. J., "Accidents and Destructive Tests," The Technology of Nuclear

Reactor Safety, MIT Press, 1964.

U.S. Atomic Energy Commission, TID-14844, J. J. DiNunno, "Calculation of

Distance Factors for Power and Test Reactor Sites," March 23, 1961.

U.S. Nuclear Regulatory Commission, NUREG-0849, "Standard Review Plan for

Research Reactor Emergency Plans," 1985.

U.S. Nuclear Regulatory Commission, NUREG-0851, 'Nomograms for

Evaluation of Doses From Finite Noble Gas Clouds," 1983.

NUREG-1537. PART I 28 Rsv 0.2/96

NUREG- I 537, PART I 28 REV 0. 2196
HEU TO LEU CONVERSIONS

U.S. Nuclear'Regulatory Commission, NUREG-1281, 'Evaluation of the

Qualification of SPERT Fuel for Use in Non-Power Reactors," 1987.

U.S. Nuclear Regulatory Commission, NUREG-1282, 'Safety Evaluation Report

on High-Uranium Content, Low-Enriched Uranium-Zirconium Hydride Fuels for

TRIGA Reactors," 1987.

U.S. Nuclear Regulatory Commission, NUREG-1313, "Safety Evaluation Repori

Related to the Evaluation of Low-Enriched Uranium Silicide-Aluminum

Dispersion Fuel for Use in Non-Power Reactors, 1988.

U.S. Nuclear Regulatory Commission, NUREG-1478,"Non-Power Reactor

Operator Licensing Examiner Standards," 1993.

U.S. Nuclear Regulatory Commission, NUTREG/CR-2079,"Analysis of Credible

Accidents for Argonaut Reactors," April 1981.

U.S. Nuclear Regulatory Commission, NUREG/CR-2198, "Fuel Temperatures in

an Argonaut Reactor Core Following a Hypothetical Design Basis Accident

(DBA)," June 1981.

U.S. Nuclear Regulatory Commission, NUREG/CR-2387, "Credible Accident

Analyses for TRIGA and TRIGA-fueled Reactors," April 1982.

U.S. Nuclear Regulatory Commission, NUREG/CR-4203, "A Calculational

Method for Determining Biological Dose Rates From Irradiated Research Reactor

Fuel," April 1985.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.25, 'Assumptions Used

for Evaluating the Potential Radiological Consequences of a Fuel Handling

Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized

Water Reactors," March 1972.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.109, "Calculation of

Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose

of Evaluating Compliance With 10 CFRPart 50, Appendix I," October 1977.

U.S. Nuclear Regulatory Commission, Regulatbry Guide 2.6, "Emergency

Planning for Research and Test Reactors," March 1983.

U.S. Nuclear Regulatory Commission, DG8005, 'Assessing External Radiation

Doses From Airborne Radioactive Materials," 1991.

REV. 0.2/96 STANDARD FORMAT n CoNTErr

REV. 0,296 - -29- STAimARD FoRmAT AND CoNTEN-T
MAPPENDVC 18.1

U.S. Nuclear Regulatory Commission, Memorandum to R. L. Tedesco from L. S.

Rubenstein, "Design Basis Event for The University of Michigan Reactor," June

17, 1981.

Woodruff W. L., 'A Kinetics and Thermal-Hydraulics Capability for the Analysis

of Research Reactors," Nuclei Technology, 64, pp. 196-206, February 1984.

VNUREG-1537, PART I 30 REV 0,2/96

NUREG 1537,PARTI 30 REV 0 2/96
APPENDIX A

APPLICABILITY OF SELECTED REGULATIONS

IN TITLE 10, CHAPTER I,

OF THE CODE OF FEDERAL REGULATIONS

TO NON-POWER REACTORS

STANDARD FoRMAT AND COMN rT
Applicability of Selected Regulations in Title 10,

Chapter I, of the Code of Federal Regulations

to Non-Power Reactors

Licensees have occasionally asked the staffabout the applicability of certainregulations to non-power reactors. In this appendix to the format and content

guide, the staff gives the results of its assessment of the applicability of the

regulations in selected parts of Title 10 of the Code of Federal Regulations (10

CFR), Chapter 1, to non-power reactors. The assessment was based on the staffs

technical background and past regulatory experience. This appendix is not an

interpretation of the regulations and should not be used by licensees and other

interested parties as such. The applicability of the regulations to non-power

reactors as determined by the assessment will not be recognized to be binding on

the Commission. Except as specifically authorized by the Commission in writing,

no interpretation of the meaning of the regulations by any officer or employee of

the Commission other than a written interpretation by the General Counsel will be

recognized to be binding on the Commission.

The regulations are updated as the need arises by rulemaking. This appendix is

based on the regulations that were in effect as of July 1, 1995. In assembling the

information in this appendix, the staff assumed that there are no commercial (as

defined by 10 CFR 50.22) non-power reactors, although for the purpose of fee

collection, there are several commercial non-power reactors as defined in 10 CFR

Part 170, which uses a different test than does 10 CFR Part 50. None of the nonpower reactors licensed today meet the 10 CFR Part 50 definition of commercial.

The applicability of the regulations for a commercial non-power reactor as defined

in 10 CFR Part 50 would be different from that for non-commercial, non-power

reactors.

The regulations are presented in two sections.

* Section L 'Selected 10 CFRParts Applicable to Non-PowerReactors

(Including Test Reactors) Licensed in Accordance With 10 CFR Part 50,"

contains a listing of all of those sections in 10 CFR Parts 19, 20, 30, 50,

70, 170, and 171 that are applicable to non-power reactors. Several

regulations in this section are only applicable to research reactors and are

clearly marked as such.

* Section UI, "Selected 10 CFR Parts Applicable to Test Reactors Licensed in

Accordance With 10 CFR Part 50," contains a listing of additional

regulations in 10 CFR Parts 20 and 50 that are applicable only to test

reactors (also called testing facilities in certain regulations).

REV. 0, 2/96 Al STANDARD FORMAT AND CONTENT

REV. O. 2/96 A-l STAEDARD FORMAT AND COwNTEN
APPENDDKA

Other parts of the regulations may also be applicable to non-power reactors if

certain events occur. For example, 10 CFR Part 36, "Licenses and Radiation J

Safety Requiremnents for Irradiators," may be applicable to non-power reactors if

they have an irradiator. Because certain events must occur for other parts of the

regulations to be applicable, the staff cannot make a statement about the

applicability of other parts to all non-power reactors.

NUREG-1537,PT I A.2 REv. 0,2196 NUREG- 1 537. PART I A-2 REV. O. 2/96
AppucABIUT oF sELECTE REGuLATioNs

SECTION I

SELECTED 10 CFR PARTS APPLICABLE TO

NON-POWER REACTORS (INCLUDING TEST

REACTORS) LICENSED IN ACCORDANCE

WITH 10 CFR PART 50

Part 19 - Notices, Instructions, and Reports to Workers.:

Inspection and Investigations

19.1 Purpose

19.2 Scope

19.3 Definitions.

19.4 Interpretations

19.5 Communications

19.8 Information collection requirements: 0MB [Office of

Management and Budget] approval

19.11 Posting of notices to workers

19.12 ~ Instructions to workers

19.13 Notifications and reports to bidividuals

19.14 Presence of representatives of licensees and workers during

inspections

19.15 Consultation with workers during inspections

19.16 Requests by workers for inspections

19.17 Inspections not warranted; informal review

19.18 Sequestration of witnesses and exclusion of counsel in

interviews conducted under subpoena

19.20 Employee protection

19.30 Violations

19.31 Application for exemptions

19.32 Discrimination prouibited

19.40 Cri iiina penalties

REV. 0,2196 - A-3 - STID FouAT Contwr

REV.0O.2/96 -- A-3 - - - - -- STA,.kzARD FoRmAT mm CoNTENT -
JILAPWENDx A

Part 20 - Standards for Protection Against Radiation

Subpart A - General Provisions

20.1001 Purpose

20.1002 Scope

20.1003 Definitions

20.1004 Units of radiation dose

20.1005 Units of radioactivity

20.1006 Interpretations

20.1007 Commun cations

20.1008 Implementation

20.1009 Information collection requirements: OMB approval

Subpart B - Radiation Protection Programs

20.1101 Radiation protection programs

Subpart C - Occupational Dose Limits

20.1201 Occupational dose limits for adults.

20.1202 Compliance with requirements for surnmation of external

and internal doses

20.1203 Determination of external dose from airborne radioactive

mateal

20.1204 Determination of internal exposure

20.1205 [Reserved]

20.1206 Planned special exposures

20.1207 Occupational dose limits for minors

20.1208 Dose to an embryo/fetus

Subpart D - Radiation Dose Limits for Individual Members of the Public

20.1301 Dose limits for individual members of the public

20.1302 Compliance with dose limits for individual members of the

public

Subpart E - [Reserved]

Subpart F - Surveys and Monitoring

20.1501 General

20.1502 Conditions requiring individual monitoring of external and

internal occupational dose

NUREG-1537, PART I 44 REv 0,2/96 NUREG- 1537, PART I A-4 REV 0, 2/96
APPL-cABILT oFsLEcE:' REaoUA77oxs

Subpart G - Control of Exposure From External Sources in Restricted Areas

20.1601 Controlof access to high radiation areas

20.1602 Control of access to very high radiation areas

Subpart H -Respiratory Protection and Controls To Restrict Internal Exposure in

Restricted Areas

20.1701 Use of process or other engineering controls

20.1702 Use of other controls

20.1703 Use of individual respiratory protection equipment

20.1704 Further restrictions on the use of respiratory protection

equipment

Subpart I - Storage and Control of Licensed Material

20.1801 Security of stored material

20.1802 Control of material not in storage

Subpart J - Precautionary Procedures

20.1901

20.1902

20.1903

20.1904

20.1905

20.1906

Caution signs

Posting requirements

Exceptions to posting requirements

Labeling containers. -

Exemptions to labeling requirements ..

Procedures for receiving and opening packages

I -, .1 I1. . .,

1. , I . -

Subpart K - Waste Disposal

20.2001

20.2002

20.2003

20.2004(a)

20.2005

20.2006

20.2007

General requirements

Method for obtaining approval of proposed disposal

procedures

Disposal by release into sanitary sewerage

Treatment or disposal by incineration

Disposal of specific wastes

Transfer for disposal and manifests

Compliance with environmental and health protection

regulations - -

Subpart L - Records

20.2101

20.2102

General provisions

Records of radiation protection programs

REv 0.2/96 A.S STANDARD FoIuT AND CONTENT

REf0 2M6 A-5 STANDARD FoRmAT AND CoN7ENT
AAPPENDIX A

20.2103

20.2104

20.2105

20.2106

20.2107

20.2108

20.2110

Records of surveys

Determination of prior occupational dose

Records of planned special exposures

Records of individual monitoring results

Records of dose to individual members of the public

Records of waste disposal

Form of records

"Iy

Subpart M - Reports

20.2201(aX)I)

20.2201(a)(2)(i)

20.2201(a)(2)Cfi)

20.2201 (b)(1)

20.2201(b)(2)(ii)

20.2201(c)

20.2201(d)

20.2201(e)

20.2202(a)

20.2202(b)

20.2202(c)

20.2202(d)(1)

20.2202(d)(2)

20.2202(e)

20.2203(a)

20.2203(b)

20.2203(d)

20.2204

20.2205

Reports of theft or loss of licensed material

Notification of incidents

<I

Reports of exposures, radiation levels, and concentrations

of radioactive material exceeding the limits

Reports of planned special exposures

[Reserved]

Subpart N - Exemptions and Additional Requirements

20.2301 Applications for exemptions

20.2302 Additional requirements

Subpart 0 - Enforcement

20.2401 Violations

20.2402 Criminal penalties

NUREG-1537,PMT I A.6 REV 0,2/96 NUREG- 1 537, PART I A-6 REV 0, 2/96
APPUicABn=T oF sELcTI- REGUATIoNs

Appendix A to Part 20 - Protection Factors for Respirators

Appendix B to Part 20 - Annual Limits on Intake (Ails) and Derived Air

Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent

Concentrations; Concentrations for Release to Sewerage

Appendix C to Part 20- Quantities of Licensed Material Requiring Labeling

Appendix D to Part 20 - United States Nuclear Regulatory Commission Regional

Offices

Appendix E to Part 20 - [Reserved]

Appendix F to Part 20 -Requirements for Low-Level-Waste Transfer for Disposal

at Land Disposal Facilities and Manifests

Appendix G to Part 20 - Requirements for Transfers of Low-Level Radioactive

Waste Intended for Disposal at Licensed Land Disposal Facilities and Manifests

REV. O,2J96 A.7. ST.Q4DARD FOIMAT AND CONTh7

REX'. O. 2196 A-7. STANDARDFORMATRIMUON7EN'r '.
*-

APPENDDC A

Part 30 - Rules of General Applicability to Domestic

Licensing of Byproduct Material

Comment - Non-power reactor licenses issued under 10 CFR Part 50 contain

authorization to receive, possess, and use byproduct material pursuant to 10 CFR

Part 30. The Part 50 license states that the receipt, possession, and use of

byproduct materials as authorized by the license will be in accordance with the

Commission's regulations in 10 CFR Part 30 including Section 30.33.

General Provisions

30.1 Scope

30.2 Resolution of conflict

30.3 Activities requiring license

30.4 Definitions

30.5 Interpretations

30.6 Communications

Comment - Also see 10 CFR 50.4.

30.7 Employee protection

Comment - Also see 10 CFR 50.7.

30.8 Information collection requirements: OMB approval

30.9 Completeness and accuracy of information

Comment - Also see 10 CFR 50.9.

30.10 Deliberate misconduct

Comment - Also see 10 CFR 50.5.

Exemptions

30.11 Specific exemptions

30.12 Persons using byproduct material under certain Department

of Energy and Nuclear Regulatory Commission contracts

30.13 Carriers

30.14 Exempt concentrations

30.15 Certain items containing byproduct material

30.16 Resins containing scandium-46 and designed for sandconsolidation in oil wells

30.18 Exempt quantities

30.19 Self-luminous products containing tritium, krypton-85, or

promethium-147

30.20 Gas and aerosol detectors containing byproduct material

Licenses

30.31 Types of licenses

NUREG-1537. PART I A.8 REV 0.2/96

NUREG-1537,PART I A-8 REV 0, 2/96
APPLCABLIT OF SELECTED REGuLATioNs

30.32 Application for specific licenses

Comment - 10 CFR 50.31 allows an applicant to combine in

one, several applications for different kinds of licenses. The

application for a byproduct material license is normally part

of the Part 50 license application for a non-power reactor.

30.33 General requirements for issuance of specific licenses

30.34 - Terms and conditions of licenses

30.35 Financial assurance and recordkeeping for decommissioning

Comment - Also see 10 CFR 50.33 and 50.75.

30.36 Expiration and termination of licenses

Comment - Also see 10 CFR 50.82.

30.37 Application for renewal of licenses

Comment - Also see 10 CFR 50.51.

30.38 Application for amendment of licenses

Comment - Also see 10 CFR 50.90.

30.39 Commission action on applications to renew or amend

30.41 Transfer of byproduct material

Records, Inspections, Tests, and Reports

30.50 Reporting requirements

30.51 Records

30.52 Inspections

30.53 Tests

30.55 Tritium reports

Enforcement

30.61 Modification and revocation of licenses

30.62 Right to cause the withholding or recall of byproduct

material

30.63 Violations

30.64 Criminal penalties

Schedules

30.70 Schedule A - Exempt concentrations

30.71 Schedule B

30.72 Schedule C - Quantities of radioactive materials requiring

consideration of the need for an emergency plan for

responding to a release

Comment - Also see 10 CFR 50.47

REV 0.2/96 - -. -A-9 - STANDARD FORMAT jwr CowIr

REV 0, 2/96 :--- - -A-9.-- - - - STAND)ARD FORMAT AND COUNTEN
APPENDDC A

Appendix A to Part 30 - Criteria Relating to Use of Financial Tests and Parent

Company Guarantees for Providing Reasonable Assurance of Funds for

Decommissioning

Appendix B to Part 30

Appendix C to Part 30 - Criteria Relating to Use of Financial Tests and Self

Guarantees for Providing Reasonable Assurance of Funds for Decommissioning

NEJREG-1537, PART I A-b REV 0,2/96 NUREG- 1537, PART I A-10 REV 0, 2/96
APPLICABNTY OF SELECTED REGULATIONS

Part 50 - Domestic Licensing of Production and

Utilization Facilities

General Proisions

50.1

50.2

50.3

50.4(a)

50.4(b)

50.4(bXl)

50.4(bX2)

50.4(b)(2)(i)

50.4(b)(3)

50.4(bX4)

50.4(bX5)

50.4(bX5)(i)

50.4(bX5Xii)

50.4(bX5)Ciii)

50.4(c)

50.4(d)

50.4(e)

50.4(f)

50.5

50.7

50.8

50.9

Basis, purpose, and procedures applicable

Definitions

Interpretations

Written communications

Deliberate misconduct

Employee protection

Information collection requirements: OMB approval

Completeness and accuracy of information

Requirement of License, Exceptions

50.10(a)

50.10(bXl)

50.10(bX2)

50.10(bX3)

50.10(bX4)

50.10(c)

50.10(cXl)

50.10(cX2)

50.l0(cX3)

License required

Comment - This regulation allows universities to construct

without a construction permit multipurpose buildings that

will be used for the research reactor at some time in the"-

future. It applies to research reactors only. It does not

apply to test reactors.

Comment - This regulation applies to research reactors

only. It does not apply to test reactors.

Rpx. 0,2/96 A.1 1 STANDARD FORMAT AND CONTh7

REV. O. 2/96 A-1 I STANDARD FORMAT AND CoNTEN I. . . I
- . 0-

APPENDc A

50.10(d)

50.11 Exceptions and exemptions from licensing requirements

50.12 Specific exemptions

50.13 Attacks and destructive acts by enemies of the United

States; and defense activities

Classif cation and Description of Licenses

50.20 Two classes of licenses

50.21 Class 104 licenses; for medical therapy and research and

development facilities

50.21(a)

50.21(c)

50.22 Class 103 licenses; for comnnercial and industrial facilities

Comment - This regulation contains the test for determining

if a non-power reactor is a commercial or industrial facility.

50.23 Construction permits

Applicationsfor Licenses, Form, Contents, Ineligibility of Certain Applicants

50.30(a)(1)

50.30(a)(2)

50.30(a)(5)

50.30(aX6)

50.30(b)

50.30(c)

50.30(d)

50.30(e)

50.30(f)

50.31

50.32

50.33

50.33(a)

50.33(b)

50.33(c)

50.33(d)

50.33(e)

50.33(f)

50.33(h)

50.33(i)

50.33(k)

50.34(a)

50.34(a)(1)

50.34(a)(2)

Filing of applications for licenses; oath or affirmation

<-/_

[Removed]

Conbining applications

Elimination of repetition

Contents of applications; general information

Contents of applications; technical information

'x

NUREG-1537. PART 1 A- 12 REV 0.2/96 NUREG-1537, PART I A-12 RE-V 0, 2t96
APPucABnzrlY OF sELECrED REOuLATioHs

50.34(aX3)

50.34(aX4)

50.34(aX5)

50.34(a)(6)

50.34(aX7)

50.34(aXg)

50.34(aX9)

50.34(aXIO)

50.34(b) -

50.34(bXl)

50.34(b)(2)

50.34(bX2)(i)

50.34(b)(3)

50.34(bX4)

50.34(b)(5)

50.34(b)(6)

50.34(bX6)(i)

50.34(b)(6)(ii)

50.34(bX6Kvi)

50.34(bX6)(v)

50.34(bX6)(v)

50.34(b)(6)(vi)

50.34tb)(7)

50.34(b(8)

50.34(c)

50.34(d)

50.34(c)

50.35

50.36(a)

50.36(b)

50.36(c)

50.36(cXI)

50.36(c)(1)(i)(A)

50.36(cXl)(i)i)(A)

50.36(cX2)

50.36(cX3)

50.36(c)(4)

50.36(c)(5)

50.36(c)(6)

50.36(cX7)

50.36(d)

50.37

50.38

50.39

Comment - See also 10 CFR 73.60.

Isuance of construction permits

Technical specifications

Agreement limiting access to restricted data

ineligibility of certain applicants

Public inspection of applications

KEy. U, 2Ib A-13 STANDARD FORMAT CoN'rr

REV. U. ZM6 A-13 STANDARD FoRmAT AND CoNTENT
* -

APPENDDC A

Standardsfor Licenses and Construction Permits

50.40 Common standards

50.41 Additional standards for Class 104 licenses

50.41(a)

50.41(b)

50.45 Standards for construction permits

' )

Issuance, Limitations and Conditions of Licenses and Construction Permits

50.50

50.51

50.52

50.53

50.54

50.54(b)

50.54(c)

50.54(d)

50.54(e)

50.54(f)

50.54(g)

50.54(h)

50.54(i)

50.54(1-1)

50.540)

50.54(k)

50.540)

50.54(m)(1)

50.54(n)

50.54(P)

50.54(q)

50.54(r)

50.54(v)

50.54(x)

50.54(y)

50.54(aa)

50.54(cc)

50.54(dd)

50.54(ee)

50.55

50.55(a)

50.55(b)

50.55(c)

50.55(d)

Issuance of licenses and construction permits

Duration of licenses, renewal

Combining licenses

Jurisdictional limitations

Conditions of licenses

<2

Conditions of construction permits

NUREG-1537, PART I A- 14 KEV UUYtNUREG 1537,PARTI A- 14 KEV V. /Y
APVLicABwT~y oF sEL cEC: REGuLAn1oms

50.55(e)

50.55a

50.55a(aXl)

50.55a(aX3)

50.SSa(b)

S0.55a(bXl)

50.55a(bX2)

50.55a(bX2)(i)

50.55a(b)(2Xii)

50.55a(bX2)(iv)

50.55a(bX2Xiv)(A)

50.55a(b)(2)(v)

50.55a(b)(2)(vi)

50.55a(b)(2)(vii)

50.55a(b)(2Xviii)

50.55a(c)

50.55a(c)(1)

50.55a(c)(2)

50.55a(c)(3)

50.55b

50.56

50.57

50.59

50.64

Codes and standards

[Reserved]

[Revoked]

Conversion of construction permit to license; or amendment

of license

Issuance of operating ce.,

Changes, tests, and eperiments

imitations on the use of highly enriched uranium (HEU) in

domestic non-power reactors

Inspections, Records, Reports. Notifications

50.70(a)

50.70(b)(1)

50.70(bX3)

50.71(a)

50.71(c)

50.71(d)

50.74

50.75(a)

50.75(d)

50.75(eXl)

50.75(e)(2)

50.75(f)

50.75(g)

Inpections

Maintenance of records, ialding of reports

Notification of change in operator or senior operator status

Reporting and recordkeeping for decommissioning planning

_ . . . .

REV.0,2/6 . .

- A-15STADARDFORMT . . AD COTEN

REV. 0, 2/96' - -- A-15' STAmARD FoPmAT Am CoNrENT
I AAPPEmix A .--

USIIAEA [International Atomic Energy Agency] Safeguards Agreement

50.78 Installation information and verification

Transfers of License - Creditors' Rights - Surrender of Licenses

50.80

50.81

50.82(a)

50.82(b)

50.82(bXl)

50.82(bXl)(ii)

50.82(bXl)(iii)

50.82(bX2)

50.82(bX3)

50.82(bX4)

50.82(b)(5)

50.82(c)

50.82(d)

50.82(e)

50.82(f)

Transfer of licenses

Creditor regulations

Application for termination of license

Amendment of License or Construction Permit at Request of Holder

50.90 Application for amendment of license or construction permit

50.92(a) Issuance of amendment

50.92(b)

Revocation, Suspension. Modification, Amendment of Licenses add Construction

Permits, Emergency Operations by the Commission

50.100 Revocation, suspension, modification of licenses and

construction permits for cause

50.101 Retaking possession of special nuclear material

50.102 Commission order for operation after revocation

50.103 Suspension and operation in war or national emergency

Enforcement

50.110 Violations

50.111 Criminal penalties

NUREG-1537, PART I A.16 RE'. 0,2/96 NUREG- 1 537, PART I A-16 REV. O. 2/96
APPLiCAEILrY OF SELECED REGULAnoNs

Appendix E to Part SO - Emergency Planning and Preparedness for Production and

Utilization Facilities

I Introduction ' II .The preliminary safety analysis report

Ell The final safty anialysis report

IV Content of eergency plans

V Implementing procedures

. . . .

R~v.0, 196A-I7 - STA~AR FoT A COTN.

REV. 0, 21% -A-17 - - -- -STANDAm FoRmAT Am CotTrENT ..
asAPPENx A

Part 70 - Domestic Licensing of Special Nuclear Material

Comment - Non-power reactor licenses issued under 10 CFR Part 50 contain

authorization to receive, possess, and use special nuclear material pursuant to

10 CFR Part 70. The Part 50 license states that the receipt, possession, and use of

special nuclear material as authorized by the license will be in accordance with the

Commission's regulations in 10 CFR Part 70 including 10 CFR 70.23 and 70.31.

General ProWisions

70.1(a) Purpose

70.2 Scope

70.3 License requirements

70.4 Definitions

70.5 Communications

Comment - Also see 10 CFR 50.4.

70.6 Interpretations

70.7 Employee protection

Comment - Also see 10 CFR 50.7.

70.8 Information collection requirements: OMB approval

70.9 Completeness and accuracy of information

Comment - Also see 10 CFR 50.9.

70.10 Deliberate misconduct

Comment - Also see 10 CFR 50.5.

Exemptions

70.11 Persons using special nuclear material undqr certain

Department of Energy and Nuclear Regulatory Commission

contracts

70.12 Carriers

70.14(a) Specific exemptions

70.14(b) [Reserved]

General Licenses

70.18 Types of licenses

70.19 General license for calibration or reference sources

70.20 General license to own special nuclear material

70.20a General license to possess special nuclear material for

transport

NUREG-1537, PART! A- 18 REV. 0,2196 NUREG- 1 537, PART I A-018 REV. O. 219
-Appum = OF SELECTED REGuLAnoS

License Applications

Comment - 10 CFR 50.31 allows an applicant to combine in one, several

applications for different kinds of licenses. The application for a special nuclear

material license is normally part of the Part 50 license application for a non-poWer

reactor.

70.21(aX2)

70.21(aX3)

70.21(b)

70.21(c)

70.21(d)

70.21(e)

70.21(f)

70.21(g)

70.22(a)

70.22(c)

70.22(d)

70.22(e)

70.22(g)

70.22(i

70.22(k)

70.22(1)

70.22(m)

70.23(a)

70.23(aXl)

70.23(aX2)

70.23(a)(3)

70.23(aX4)

70.23(aX5)

70.23(a)(6)

70.23(aX7)

70.23(a)(9)

70.24

70.25

Filingiz

Contents of applications

[Reserved]

Requirements for the approval of applications

Criticality accident requirements

Comment - I0 CFR 70.24(c) refers to critical assembly

reactors.

Financial assurance and recordkeeping for decommissioning

Comment'- Also see 10 CFR 50.33 and 50.75.

Licenses

70.31(a)

70.31(b)

70.31(c)

Issuance of licenses

[Deleted]

REV. O. 2/96 A-19. STANDARD FORMAT AND CoNTEr
I -

APPENDDC A

70.31(d)

70.32(a)

70.32(b)

70.32(cX2)

70.32(d)

70.32(e)

70.32(f)

70.32(g)

70.32(h)

70.320()

70.320)

70.33

70.34

70.35

70.36

70.37

70.38

Conditions of licenses

[Reserved]

[Reserved]

Renewal of licenses

Comment - Also see 10 CFR 50.51.

Amendment of licenses

Comment - Also see 10 CFR 50.90.

Commission action on applications to renew or amend

Inalienability of licenses

Disclaimer of warranties

Expiration and termination of licenses

Comment - Also see 10 CFR 50.82.

Jr

Acquisition, Use, and Transfer of Special Nuclear Materil, Creditors' Rights

70.41 Authorized use of special nuclear material

70.42 Transfer of special nuclear material

70.43 [Deleted]

70.44 Creditor regulations

Special Nuclear Material Control, Records, Reports and Inspecioens

70.50

70.51(a)

70.51(b)

70.51(c)

70.51(d)

70.51(i)

70.52

70.53(a)

70.54

70.55(a)

Reporting requirements

Material balance, inventory, and records requirements

Reports of accidental criticality or loss or theft or attempted

theft of special nuclear material

Material status reports

Nuclear material transfer reports

Inspections

Comment - Also see 10 CFR 50.70.

70.55(b)

70.55(cX3)

NUREG.1537, PART I A-20 REv. 0, 2/96

NUREG- 1537, PART I A-20 RE.0, 296
AppucAmLnBy OF Sg LCTED REGuLATnoNs

70.56 .Tests70.60 [Removed] ,

Modificaion andRevocation of Licenses

70.61 - Modification and revocation of licenses

70.62 Suspension and operation in war or national emergency

Enforcement

70.71 Violations

70.72 Criminal penalties

REV. 0,2/96 A 21 I STANDARD FORMAT AmW CONTENT
I -

APPENDv A

Part 170 - Fees for Facilities, Materials, Import and

Export Licenses and Other Regulatory Services Under

the Atomic Energy Act of 1954, as Amended

Comment - This part applies to those non-power reactor licensees that are subject

to licensing fees (application fees and hourly fees).

General Provisions

111 )

170.1

170.2

170.3

170.4

170.5

170.8

170.11

170.12(a)

170.12(b)(2)

170.12(c)(2)

170.12(d)(2)

170.12(eXl)

170.12(g)

170.12(h)

170.12(i)

170.20

Purpose

Scope

Definitions

Interpretations

Communications

Information collection requirements: OMB approval

Exemptions

Comment -This section contains the exemption for

nonprofit educational institutions, Government agencies,

and State-owned research reactors used primarily for

educational training and academic research purposes.

Payment of fees

Average cost per professional staff-hour

Schedule of Fees

170.21 Schedule of fees for production and utilization facilities,

review of standard referenced design approvals, special

projects, inspections, and import and export licenses

Schedule of fees for materials licenses and other regulatory

services, including inspections, and import and export

licenses

170.31

Enforcement

170.41

170.51

Failure by applicant or licensee to pay prescribed fees

Right to review and appeal of prescribed fees

NURBG-1S37,PT I A-22 REV.O,2196 NUREG-1537, PART I A-22 R.EV. 0, 2/96
APPLCABnLtmY OF sELEcFED REGuLAnoNs

Part 171 - Annual Fees for Reactor Operating Licenses,

and Iwdilj;Hdmim3')Ms5!et L42ompkance,

Registrations, and Quality Assurance Program Approvals

and Government Agencies Licensed by NRC

Comment - This part applies to those non-power reactor licensees that are subject

to annual fees.

171.1 Purpose

171.3 Scope

171.5 Definitions

171.7 Interpretations

171.8 Information collection requirements: 0MB approval

171.9 Communications

171.11(a) Exemptions

Comment - This section contains exemptions for nonprofit

educational institutions under some circumstances and for

federally and State-owned research reactors used primarily

for educational training and academic research purposes.

171.11(b)

171.11(c)

171.13 Notice

171.15(a) Annual fees: reactor operating licenses

171.15(e)

171.15(f)

171.16 Annual fees: material licensees, holders of certificates of

compliance, holders of sealed source and dVice

registrations, holders of quality assurance program

approvals, and government agencies licensed by the NRC

171.17 Proration

171.17(a)

171.19 Payment

171.23 Enforcement

171.25 Collection, interest, penalties, and administrative costs

REv. 0,2/96 A-23 STANDARD FORMAT AND CoNImT

REX',. 0, 216 A-23 STANDAm FoRmAT AND CoNm-,r
- ApFLIcABInr oF saE=E REuuLAToN

-- -SECTION II

SELECTED 10 CFR PARTS APPLICABLE TO

TEST REACTORS LICENSED

IN ACCORDANCE WITE 10 CFR PART 50

Part 20 - Standards for Protection Against Radiation

Subpart M - Reports

20.2206 Reports of individual moriitOring

.. ... ,

R~v.0,296 A25.STANARDFoRMT AD COTE_

REV. 0, 2M A-25'. ,-. STANDARD FoRmAT Am Comarr
aAPPENDDC A

Part 50 - Domestic Licensing of Production and

Utilization Facilities

Requirement of License, Exceptions

50.10(e) License required

Applicationsfor Licenses, Form, Contents, Ineligibility of Certain Applicants

50.30(aX3) Filing of applications for licenses; oath or affirmation

50.30(aX4)

50.36b Environmental conditions

Issuance, Limitations, and Conditions of Licenses and Construction Permits

50.58 Hearings and report of the Advisory Committee on Reactor

Safeguards

Inspections, Records, Reports. Notifications

50.71(b) Maintenance of records, maldng of reports

Amendment of License or Construction Permit at Request ofHolder

50.91 Notice for public comment; State consultation

50.91(a)

50.91(b)

50.91(cXl) "

50.92(c) Issuance of amendment

Appendbc C to Part 50 - A Guide for the Financial Data and Related Information

Required To Establish Financial Qualifications for Facility Construction Permits

Appendix Q to Part 50 - Pre-Application Early Review of Site Suitability Issues

NUREG-1537, PART! A-26 REv. 0. 2/96

NUREG- 1537, PART I A-26 REV. 0. 2/96
I NRC FORU 335S

NRCM 1102

3201 320

US NUCLEAR REGULATORY COMMISSION

BIBLIOGRAPHIC DATA SHEET

1see ImftlJCTons on the reerse)

*TLE AND SUBTITLE

Guidelines for Preparing and Reviewing Applications for

the Licensing of Non-Power Reactors. Part 1, Format and

Content.

1 RFPORT NUMBER

(Angned bYf RC Add Vol. S&upp Pw.r att Adfdum No bo m-? I

NUREG-1537

Part 1

3 DATE REPORT PUBLISHED

MONTH YEAR

February 1996

4 FIN OR GRANT NUMBER

S AUTHOR(S) 6 TYPE OF REPORT

7 PERIOD COVERED uIirws.* Dtes)

_ ______I 8 PERFORMING ORGANIZATION-NAME AND ADDRESS III NRC prrvMde Devwon 011w. etrRapo. .US NwebLr C'.?nin, .d AS As. I? entr Dank p, 

,- mod w - YddlJ

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

9 SPONSORING ORGANIZATION-NAME AND ADDRESS 1if NAC pe meaS cb .O *1fCntretlo ronP de NRCD- n Of 1 or Rfeov, US Ni R.,Aepiory CommE.n

-nd - a. J 

Same as above

-UPPLEMENTARY NOTES

BSTRACT 1200 o- oeti

NUJREG-1537, Part 1 gives guidance to non-power reactor licensees and

applicants on the format and content of applications to the Nuclear

Regulatory Comnission for licensing actions. These licensing actions

include construction permits and initial operating licenses, license renewals,

amendments, conversions from highly enriched uranium to lQw-enriched uranium,

decommissioning, and license termination. %

12 KEYWDRDSbESCRPTORSLuttr

Non-Power Reactor

Research Reactor

Test Reactor

Format and Content Guide

13 AVAILABILITY STATEMENT

Unlimited

14 SECURITY CLASSIFICA71ION

|T{ts Paw)

Unclassified

I rho 'Repolt)

Unclassified

1[ NUMBER OF PAGES

16 PRICE

NRC FORM 13S 12 891
Mb,.

Federal Recycling Program