NUREG-1022

Rev. 1

Event Reporting Guidelines

10 CFR 50.72 and 50.73

U.S. Nuclear Regulatory Commission

Office for Analysis and Evaluation of Operational Data
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Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.

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

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10018-3308.
NUREG-1022

Rev. 1

Event Reporting Guidelines

10 CFR 50.72 and 50.73

Manuscript Completed: December 1997

Date Published: January 1998

Prepared by

D. P. Allison, M. R Harper, W. R. Jones,

J. B. MacKinnon, S. S. Sandin

Office for Analysis and Evaluation of Operational Data

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

,* 0
Impact

This Revision 1 to NUREG-1022 clarifies and consolidates the guidance on implementing the

event notification and reporting requirements in 10 CFR 50.72 and 50.73. Little of the guidance

is new or different from the generic reporting guidance previously published in final form in

NUREG-1 022 (1983), its Supplement 1 (1984) and subsequent generic communications.

Where it is different, the changes are minor. In some areas the new guidance will result in

fewer reports and in some areas it will result in more reports. On balance, the clarified

guidance will result in a small decrease in reporting burden.

PAPERWORK REDUCTION ACT STATEMENT

This report amends the information collections contained in 10 Code of Federal Regulations

(CFR) Part 50 and NRC Form 366, Licensee Events Reports. The changes are considered to

be insignificant when compared with the overall requirements of the CFR Part and the form

(NRC Form 366 reduction of 350 hours annually vs. the current 75K, and 10 CFR 50.72

reduction of 150 hours annually vs. the current 2.4K). NRC does not consider the burden to be

significant enough to trigger the requirements of the Paperwork Reduction Act of 1995

(44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management

and Budget (OMB), approval number 3150-0011 and 3150-0104.

Public Protection Notification

If an information collection does not display a currently valid OMB control number, the NRC

may not conduct or sponsor, and a person is not required to respond to, the information

collection.

NUREG-1022, Rev. 1 ii
ABSTRACT

Revision I to NUREG-1022 clarifies the immediate notification requirements of Title 10 of the

Code of Federal Regulations, Part 50, Section 50.72 (10 CFR 50.72), and the 30-day written

licensee event report (LER) requirements of 10 CFR 50.73 for nuclear power plants. This

revision was initiated to improve the reporting guidelines related to 10 CFR 50.72 and 50.73

and to consolidate these guidelines into a single reference document. A first draft of this

document was noticed for public comment in the Federal Register on October 7, 1991 (56 FR

50598). A second draft was noticed for comment in the Federal Register on February 7, 1994

(59 FR 5614). This document updates and supersedes NUREG-1022 and its Supplements 1

and 2 (published in September 1983, February 1984, and September 1985, respectively). It

does not change the reporting requirements of 10 CFR 50.72 and 50.73.

iii NUREG-1022, Rev. 1 -

CONTENTS

PAPERWORK REDUCTION ACT STATEMENT ................................... ii

A BSTRA C T ............................................................... iii

EXECUTIVE SUMMARY ...................................................... ix

ACKNOW LEDGMENTS ...................................................... x

ABBREVIATIO NS .......................................................... xi

I INTRO DUCTIO N ......................................................... 1

1.1 Background ..................................................... 1

1.2 Reporting Guidelines and Industry Experience ........................... 2

1.3 Revised Reporting Guidelines ....................................... 2

1.4 How to Use These Guidelines ............... ....................... 5

1.5 New or Different Guidance .......................................... 6

1.6 Planned Future Actions ............................................ 6

2 REPORTING AREAS WARRANTING SPECIAL MENTION ........................ 13

2.1 Engineering Judgment ............................................ 13

2.2 Differences in Tense Between 10 CFR 50.72 and 50.73 .................. 13

2.3 Multiple Failures and Related Events ................................. 13

2.4 Deficiencies Discovered During Engineering Reviews .................... 14

2.5 Engineered Safety Features Actuations ............................... 14

2.6 Events Discussed with the NRC Staff ................................. 14

2.7 Multiple Component Failures ....................................... 15

2.8 Preparation of Licensee Event Reports ............................... 15

2.9 Voluntary Reporting .............................................. 15

2.10 Retraction/Cancellation of Event Reports .............................. 16

2.11 Time Limits for Reporting .......................................... 16

2.12 Outside Design Basis ............................................. 17

3 SPECIFIC REPORTING GUIDELINES ........................................ 19

3.1 Section 50.72 and 50.73 General Requirements ......................... 20

3.1.1 Immediate Notification Requirements ........................... 20

3.1.2 Licensee Event Report System ............................... 22

3.2 One-hour ENS Notifications and 30-Day LERs ........................... 23

3.2.1 Plant Shutdown Required by Technical Specifications .............. 24

3.2.2 Technical Specification Prohibited Operation or Condition ........... 27

3.2.3 Technical Specification Deviation per §50.54(x) ................... 33

V NUREG-1022, Rev. I
3.2.4 Operating Plant Found in Degraded or Unanalyzed Condition ........ 34

3.2.5 External Threat to Plant Safety ............................... 41

3.2.6 ECCS Discharge Into the Reactor Coolant System ................ 45

3.2.7 Loss of Emergency Preparedness Capabilities ................... 47

3.2.8 Internal Threat to Plant Safety ................................ 51

3.3 Four-hour ENS Notifications and LERs ................................. 54

3.3.1 Shutdown Plant Found in Degraded or Unanalyzed Condition ........ 55

3.3.2 Actuation of an Engineered Safety Feature or the RPS ............. 57

3.3.3 Event or Condition That Alone Could Prevent Fulfillment of a Safety

Function ....... ................................... 65

3.3.4 Common-cause Failures of Independent Trains or Channels ......... 80

3.3.5 Airborne or Liquid Effluent Release ............................ 84

3.3.6 Contaminated Person Requiring Transport Offsite ................. 87

3.3.7 News Release or Other Government Notifications ................. 88

3.3.8 Spent Fuel Storage Cask Notifications .......................... 92

3.4 Followup Notification ............................................... 93

4 EMERGENCY NOTIFICATION SYSTEM REPORTING ........................... 95

4.1 Emergency Notification System ...................................... 95

4.2 General ENS Notification ........................................... 96

4.2.1 Tim eliness ............................................... 96

4.2.2 Voluntary Notifications ...................................... 96

4.2.3 ENS Notification Retraction .................................. 96

4.2.4 ENS Event Notification Worksheet (NRC Form 361) ............... 96

4.3 Typical ENS Reporting Issues ....................................... 97

5 LICENSEE EVENT REPORTS .............................................. 99

5.1 LER Reporting Guidelines .......................................... 99

5.1.1 Submission of LERs .................................. 99

5.1.2 LER Forwarding Letter and Cancellations ...................... 99

5.1.3 Report Legibility ......................................... 100

5.1.4 Exem ptions ............................................ 100

5.1.5 Voluntary LERs ......................................... 100

5.1.6 Supplemental Information and Revised LERs .................. 101

5.1.7 Special Reports ......................................... 101

5.1.8 Appendix J Reports (Containment Leak Rate Test Reports) ....... 102

5.1.9 10 CFR Part 21 Reports ................................... 102

5.1.10 Section 73.71 Reports .................................... 103

5.1.11 Availability of LER Forms .................................. 103

5.2 LER Content Requirements and Preparation Guidance ................... 103

5.2.1 Narrative Description or Text (NRC Form 366A, Item 17) ......... 105

5.2.2 Abstract (NRC Form 366, Item 16) ........................... 114

5.2.3 Other Fields on the LER Form .............................. 114

NUREG-1022, Rev. 1 vi
APPENDICES

A Historical Perspective on Event Reporting

B Emergency Notification System Process

C Licensee Event Report Review Programs

D 10 CFR 50.72 Including Statement of Considerations

E 10 CFR 50.73 Including Statement of Considerations

F 1992 Revision to 10 CFR 50.72 and 50.73 Including Statement of Considerations

TABLES

1. Comparability of 10 CFR 50.72 and 50.73 Criteria ................................ 7

2. Exam ple System s ....................................................... 64

vii NUREG-1022, Rev. I

EXECUTIVE SUMMARY

Two of the many elements contributing to the safety of nuclear power are emergency response

and the feedback of operating experience into plant operations. These are achieved partly by

the licensee event reporting requirements of Title 10 of the Code of Federal Regulations, Part

50, Sections 50.72 and 50.73 (10 CFR 50.72 and 50.73), which became effective on January 1,

1984. Section 50.72 provides for immediate notification requirements via the emergency

notification system (ENS) and Section 50.73 provides for 30-day written licensee event reports

(LERs).

The information reported under 10 CFR 50.72 and 50.73 is used by the NRC staff in responding

to emergencies, monitoring ongoing events, confirming licensing bases, studying potentially

generic safety problems, assessing trends and patterns of operational experience, monitoring

performance, identifying precursors of more significant events, and providing operational

experience to the industry.

Experience has shown that the threshold for reporting has not been consistently implemented

and some problems exist with the interpretation of the guidelines and definitions. A 1990

survey on the effect of NRC regulation on nuclear power plant activities and subsequent event

reporting workshops also indicated a need for further guidance on the two reporting rules.

Therefore, the NRC staff prepared NUREG-1022, Revision 1, which clarifies implementation of

the existing 10 CFR 50.72 and 50.73 rules and consolidates important NRC reporting guidelines

into one reference document. The clarifications include major editing of the previous

guidelines. The document is structured to assist licensees in achieving prompt and complete

reporting of specified events and conditions. The revised guidelines are not expected to result

in a significant change in the annual industry-wide total numbers for ENS notifications and

LERs. The effect on individual licensees is expected to vary.

The document addresses general issues of reporting that have not been consistently applied

and covers such diverse subjects as engineering judgment, multiple failures and related events,

deficiencies discovered during licensee engineering reviews, and human performance issues.

The guidelines for specific reporting criteria have been enhanced by improved discussions of

concepts, thresholds, and illustrative examples; definitions of key terms and phrases; and

original ENS guidelines for some criteria that were not previously addressed. A new section

has been added that discusses ENS communications and methods, voluntary reporting,

retraction of reports, importance of reporting timeliness and completeness, and typical NRC

concerns associated with ENS notifications for each reporting requirement.

ix NUREG-1022, Rev. 1
ACKNOWLEDGMENTS

The authors Revision I wish to acknowledge the members of the task group that prepared the

first draft of Revision 1: John Boardman, Paul Bobe, Marcel Harper, John Crooks, L. Mark

Padovan, and Robert Spence of the U.S. Nuclear Regulatory Commission (NRC) Office for

Evaluation and Operation of Operational Data (AEOD). They were assisted by Roger Woodruff

of the NRC Office of Nuclear Reactor Regulation (NRR) and Eric Weiss and Aimee Brown of

AEOD. With regard to the second draft and this final version, the authors wish to acknowledge

the contributions of Sanford Israel and Shirley Rohrer of AEOD. With regard to the final draft

the authors also wish to acknowledge the assistance of David Skeen of NRR, Milton Vagins of

the NRC Office of Nuclear Regulatory Research, and representatives of the NRC's four regional

offices, Gordon Hunegs (RI), Charles Ogle (RiI), Edward Schweibinz (RIII) and Michael Runyan

(RIV). Dee Gable of the NRC Office of Administration (ADM) reviewed the first two drafts.

Geary Mizuno of the NRC Office of the General Counsel (OGC) reviewed both drafts and the

final version.

NUREG-1022, Rev. I x
ABBREVIATIONS

AEOD Analysis and Evaluation of Operational Data, Office for

AIT augmented inspection team

ASME American Society of Mechanical Engineers

ASP accident sequence precursor

ATWS anticipated transient without scram

BPV Boiler and Pressure Vessel Code (ASME)

BWR boiling-water reactor

CFR Code of Federal Regulations

CRDM control rod drive mechanism

CRVS control room ventilation system

DBDR design-basis documentation review

DDR design document reconstitution

ECCS emergency core cooling system

EDG emergency diesel generator

EIIS Energy Industry Identification System

ENS emergency notification system

EO emergency officer

EOF emergency operations facility

EOP emergency operating procedure

EPIX equipment performance and information exchange

EPA Environmental Protection Agency (U.S.)

ERDS emergency response data system

ERF emergency response facility

ESF engineered safety feature(s)

ESW emergency service water

FEMA Federal Emergency Management Agency

FFD fitness for duty

FSAR final safety analysis report

FTS federal telecommunications system

GDC general design criteria

GL generic letter

HOO headquarters operations officer

xi NUREG-1022, Rev. I
HP health physics

HPCI high-pressure coolant injection

HPI high-pressure injection

HPN health physics network

HPSI high pressure safety injection

HVAC heating, ventilation and air conditioning

IEEE Institute of Electrical and Electronics Engineers

lIT incident investigation team

ILRT integrated leak rate test

IN information notice

INPO Institute of Nuclear Power Operations

ISI inservice inspection

IST inservice testing

ISTS improved standard technical specifications

LCO limiting condition for operation

LER licensee event report

LOCA loss of coolant accident

LPSW low pressure service water

MPC maximum permissible concentration

MSIV main steam isolation valve

NPRDS nuclear plant reliability data system

NRC Nuclear Regulatory Commission (U.S.)

NRR Nuclear Reactor Regulation, Office of

NUMARC Nuclear Management and Resources Council

OCR optical character reader

OMS overpressure mitigation system

PDR Public Document Room

PGA policies, guidance, and administrative controls

PWR pressurized water reactor

RAB Reactor Analysis Branch

RBVS reactor building ventilation system

RCIC reactor core isolation cooling

RCP reactor coolant pump

RCS reactor coolant system

RDO regional duty officer

RHR residual heat removal

RPS reactor protection system

RWCU reactor water cleanup

SALP systematic assessment of licensee performance

NUREG-1022, Rev. 1 xii
SAR safety analysis report

S/D shutdown

SIS safety injection system

SOV solenoid-operated valve

SPDS safety parameter display system

SRO senior reactor operator

STS standard technical specifications

TS technical specification(s)

TSC technical support center

°°°i NUREG-1022, Rev. 1

I INTRODUCTION

This document provides guidance on the reporting requirements of Title 10 of the Code of

Federal Regulations, Part 50, Sections 50.72 and 50.73 (10 CFR 50.72 and 10 CFR 50.73).

While these reporting requirements range from immediate, 1-hour, and 4-hour verbal

notifications to 30-day written reports, covering a broad spectrum of events from emergencies

to generic component level deficiencies, the NRC wishes to emphasize that reporting

requirements should not interfere with ensuring the safe operation of a nuclear power plant.

Licensees' immediate attention must always be given to operational safety concerns.

1.1 Background

The origins of 10 CFR 50.72 and 50.73 are described in Appendix A to this report. In 1983,

partially in response to lessons from the Three Mile Island accident, the U.S. Nuclear

Regulatory Commission (NRC) revised its immediate notification requirements via the

emergency notification system (ENS) in 10 CFR 50.72 and modified and codified its written

licensee event report (LER) system requirements in 10 CFR 50.73. The revision of 10 CFR

50.72 and the new 10 CFR 50.73 became effective on January 1, 1984. Together, they specify

the types of events and conditions reportable to the NRC for emergency response and

identifying plant-specific and generic safety issues.

The two rules have identical reporting thresholds and similar language whenever possible.

They are complementary and of equal importance, with necessary dissimilarities in reporting

requirements to meet their different purposes, as illustrated in this report, Section 1, Table 1,

and Section 3 text.

Section 50.72 is structured to provide telephone notification of reportable events to the NRC

Operations Center within a time frame established by the relative importance of the events.

Events are categorized as either emergencies (immediate notifications, but no later than 1 hour)

or non emergencies. The latter is further categorized into 1-hour and 4-hour notifications; nonemergency events requiring 4-hour notifications generally have slightly less urgency and safety

significance than those requiring 1-hour notifications. Immediate telephone notification to the

NRC Operations Center of declared emergencies is necessary so the Commission may

immediately respond. Reporting of non-emergency events and conditions is necessary to

permit timely NRC followup via event monitoring, special inspections, generic communications,

or resolution of public or media concerns.

Section 50.73 requires written LERs to be submitted on reportable events within 30 days of

their occurrence, after a thorough analysis of the event, its root causes, safety assessments,

and corrective actions are available, to permit NRC engineering analyses and studies.

1 NUREG-1022, Rev. 1
Some reporting guidance for 10 CFR 50.72 and 50.73 was contained in the Statements of

Considerations for the rules. More detailed guidelines and examples of reportable events were

developed and issued in NUREG-1022 and its Supplements 1 and 2. The intent of these

publications was to achieve complete reporting of specified events and conditions.

Subsequently, additional interpretations and directions on certain subjects have been issued in

NRC bulletins, information notices, and generic letters.

1.2 Reporting Guidelines and Industry Experience

Event reporting under these rules since 1984 has contributed significantly to focusing the

attention of the NRC and the nuclear industry on the lessons learned from operating experience

to improve reactor safety. In the mid-1980's, decreasing trends in the number of reactor

transients and in the number of significant events and improvements in reactor safety system

performance were noticeable. Since 1989, these trends have leveled off as fewer plants were

on a learning curve and industry completed improvements that have a high return in safety

performance. While the more obvious lessons have been extracted from operating experience,

more analyses need to be performed and new efforts need to be developed to extract further

lessons from operational data.

The operational experience submitted in accordance with 10 CFR 50.72 and 50.73 is publicly

available and has been used by other organizations in ways that are most often beneficial to

nuclear safety. However, uses in areas that were unintended, such as in prudence hearings, in

statistical presentations and comparisons of reporting rates without regard to or inclusion of a

technical analysis of the safety significance of the events, can lead to unwarranted impressions

of safety performance. In such uses, there has been a tendency to only count the number of

reported events without assessing their individual safety significance. Such misuses could

result in licensees adopting a more restrictive reporting threshold in order to reduce the number

of reportable events, although the Commission's requirement for a low threshold has not

changed. This can be counterproductive to the purpose of these rules.

Experience has shown that the threshold of reporting, as well as other areas of the reporting

rules, has not been consistently implemented. Some problems have occurred in such areas as

interpretation of the guidelines and definitions, timeliness of reporting, reporting of generic

concerns, engineering judgment, and reporting of deficiencies found during design reviews.

These problems, as well as a 1990 survey on the effect of NRC regulation on nuclear power

plant activities and subsequent event reporting workshops, identified the need for further

guidelines on the two reporting rules.

1.3 Revised Reporting Guidelines

The purpose of this revision to NUREG-1022 is to ensure events are reported as required by

improving 10 CFR 50.72 and 50.73 reporting guidelines and to consolidate these guidelines into

a single reference document. This document updates and supersedes NUREG-1022 and its

Supplements 1 and 2.

An NRC task group prepared this document principally by editing and combining the information

contained in NUREG-1022 and its Supplements 1 and 2, the Statements of Considerations for

NUREG-1022, Rev. 1 2
10 CFR 50.72 and 50.73, other NRC staff documents on event reporting (such as information

notices, bulletins, inspection manual chapters, enforcement actions, letters and memoranda),

ENS event notification reports, and LERs. A second task group prepared the second draft of

this document, principally by considering the public comments received and the requirements of

the rules, their Statements of Considerations, and previous NRC generic guidance on reporting.

In compiling this document, the information in NUREG-1022 was edited for clarity. The

paragraph-by-paragraph explanation of the LER rule, which was a restatement of guidance in

the Statement of Considerations, was preserved or more thoroughly discussed. Most of the

examples were replaced with others that have been condensed to exemplify specific reporting

thresholds.

Most of the specific questions and answers on both rules as contained in NUREG-1022,

Supplement 1, were incorporated as generic statements into the discussions or examples in

Sections 2, 3, 4, and 5 of this document. The ENS and LER rules are compared side-by-side in

Section 3.

NUREG-1 022, Supplement 2, made recommendations for improvements in LER quality;

Appendices B and D of Supplement 2 were incorporated into the discussions in Section 5.2 of

this document.

In addition, experience from responding to NRC staff and licensee inquiries in various event

reporting workshops since 1984 and ENS calls has been considered in this report. Many actual

events were summarized to exemplify event reportability in response to licensee requests. The

principal NRC staff involved in the original codification and revisions to 10 CFR 50.72 and 50.73

were consulted regarding the original intent of the regulations.

Section 2 clarifies specific areas of 10 CFR 50.72 and 50.73 that are applicable to many

reporting criteria or that historically appear to be subject to varied interpretations. It covers

such diverse subjects as engineering judgment, differences in tenses between the two rules,

retraction and voluntary reporting, legal reporting requirements, and human performance

issues.

Section 3 contains guidelines on event reporting for specific criteria in both rules by means of

discussions and examples of reported events. To minimize repetition, similar criteria from both

rules are addressed together. The format follows the order of 10 CFR 50.72 with 50.73

appropriately interwoven.

Section 3.1 addresses general methods of ENS reporting for declared emergencies and nonemergencies. Practical guidelines are given on making ENS emergency notifications.

Requirements for LER reporting regardless of plant mode, power level, or the significance of an

initiating item are specified.

Section 3.2 addresses ENS 1-hour reporting criteria and 30-day LERs. The existing ENS and

LER guidelines related to plant shutdowns required by technical specification (TS), TS

deviations per §50.54(x), and TS prohibited operations or conditions are reiterated. Plant

operation in a degraded or unanalyzed condition, or outside the plant's operating and

3 NUREG-1022, Rev. I
emergency procedures, is clarified by definitions and examples. The timing of ENS reporting of

anticipated natural phenomenon or conditions threatening plant safety is explained to ensure

good communication between licensees and the NRC during developing situations. Valid

emergency core cooling system (ECCS) discharges into the reactor coolant system are defined

and invalid ECCS discharges are identified as reportable within 4 hours as an engineered

safety feature (ESF) actuation. Additional guidelines and thresholds are given on the ENS

reporting of the loss of emergency assessment, response, or communications. The intent of

the reporting criteria on internal plant safety threats, including such examples as fire, toxic gas,

or radiation releases, is explained to also include any other internal safety threat. Floods and

spills are discussed as another typical threat to plant safety and the terms "threat" and

"significant hampering of site personnel" are defined.

Section 3.3 addresses 4-hour ENS notifications and 30-day LERs. Examples are provided for

degraded or unanalyzed conditions found while the plant is shut down. Engineered safety

feature and reactor protection systems actuations are discussed. Anticipated transient without

scram (ATWS) system actuations are addressed. The 1992 revisions to 10 CFR 50.72 and

50.73 that reduced the reporting of engineered safety features actuations are also discussed.

Terms are defined regarding the reporting criteria for events or conditions that alone could have

prevented fulfillment of the safety function required for shutdown of the reactor, removal of

residual heat, release of radioactive material, or mitigation of the consequences of an accident.

Single, common-mode, and multiple independent failures reportable under this criterion are

discussed. The discussion of LER reporting of common-mode failures of independent safety

system trains defines a number of terms and notes their importance as precursors. The

existing ENS and LER guidelines related to airborne or liquid releases are restated. Guidelines

are clarified on ENS reporting of a contaminated person requiring transport to an offsite medical

facility. The basis and report timing for the ENS reporting criteria regarding news releases or

other government notifications are explained, as necessary, so that the NRC can appropriately

respond to media or government inquiries and thresholds for reporting are clarified. The

recently issued ENS reporting criterion regarding spent fuel storage cask problems is included.

Section 3.4 addresses the requirements for immediate ENS followup notifications during the

course of an event. The requirement, means, and methods to maintain continuous or periodic

communication with the NRC during events, if so requested, are explained.

Section 4 explains ENS communications (from existing information notices), reporting

timeliness and completeness, voluntary notifications, and retractions. Appropriate ENS

emergency notification methods are described.

Section 5 reiterates previous guidelines on administrative requirements, preparation, and

submittal of LERs. It specifies the information an LER should contain and provides steps to be

followed in preparing an LER. It also includes an expanded human performance discussion to

achieve ENS and LER content that examines both equipment and human performance.

Appendix A provides the history of 10 CFR 50.72 and 50.73, associated NRC workshops, and

an NRC regulatory impact study, which was one of the factors leading to this document.

NUREG-1022, Rev. 1 4
Appendix B discusses the key NRC ENS personnel, range of NRC responses to ENS

notifications, and NRC event review.

Appendix C addresses the NRC LER analysis and evaluation programs and other uses of LERs

nationally and internationally.

Appendix D contains 10 CFR 50.72 including its Statement of Considerations as published in

the Federal Register.

Appendix E contains 10 CFR 50.73 including its Statement of Considerations as published in

the Federal Register.

Appendix F contains 1992 revisions to 10 CFR 50.72 and 10 CFR 50.73 including the

Statement of Considerations as published in the Federal Register.

1.4 How to Use These Guidelines

This NUREG was designed primarily as a reference to help licensees determine event

reportability, make ENS notifications, and prepare and submit LERs.

0 Reportability Determination

The applicable 10 CFR 50.72 and 50.73 reporting criteria are identified in the Table of

Contents of this report, as well as in the respective rules. Because these rules have

overlapping reporting requirements, it is not unusual to find an event reportable under

more than one criterion. A reportable event is to be reported under the most immediate

reporting requirements.

Generally, many events and conditions that require an ENS notification also require the

submittal of an LER, as reflected by many of the rules' parallel reporting requirements.

The reporting determination guidelines in Section 3 for both 10 CFR 50.72 and 50.73

are presented together wherever possible in the "Discussion" and "Example"

explanations for each paragraph. The differences between the ENS and LER reporting

requirements are underlined. The differences are discussed when they are important.

Key terms are defined and important concepts are identified in the "Discussion"

sections. Events used as examples may be reportable under other criteria but are

usually only evaluated for reportability under the specific criteria they appear under.

General issues, such as timeliness, can also be found in Section 2.

Other reporting requirements applicable to operating reactors include 10 CFR 50.9,

20.2202, 20.2203, 50.36, 72.74, 72.216, 73.71, and Part 21. When reports are required

under these regulations, some parts require the use of 10 CFR 50.72 and 50.73

notifications and written reports. Duplicate reporting is not required.

5 NUREG-1022, Rev. 1
0 ENS Notification

Once an event has been determined to be reportable under 10 CFR 50.72, an ENS

notification is to be made. The ENS notification time limit can be found under the

applicable §50.72 criteria in Section 3; if more than one reporting criterion applies, the

shortest time limit should be met. Guidelines on the information to be reported may be

found in Section 4.3. Practical information regarding the actual telephone call can be

found in Sections 4.1 and 4.3.

* LER Preparation and Submittal

Once an event has been determined to be reportable under 10 CFR 50.73, an LER is to

be prepared and submitted. Administrative requirements and guidelines for submitting

LERs can be found in Section 5.1. The requirements and guidelines for the content of

LERs can be found in Section 5.2.

1.5 New or Different Guidance

Reporting guidance that is considered to be new or different in a meaningful way, relative to

previously published generic reporting guidance, is indicated by shading the appropriate text.

Occasionally, text is marked by both redline and strikeout in order to show that specific items

are being deleted.

1.6 Planned Future Actions

The NRC staff recognizes a need to revise the reporting requirements in 10 CFR 50.72 and

50.73 to better align them with the NRC's current needs, including the move toward riskinformed regulation and reporting of design basis issues. Accordingly, the staff plans to

develop a rulemaking plan and proceed expeditiously with rulemaking to accomplish these

ends. However, rulemaking will take time, whereas the work on this document has been

completed. Thus, it is considered worthwhile to issue this revision to NUREG-1022 to provide

improved guidance now. In the future, as rule changes are developed, new guidance will be

developed concurrently.

NUREG-1022, Rev. I 6
Table 1 Comparability of 10 CFR 50.72 and 50.73 Criteria

-4

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0D5

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ENS notification as soon as ENS notification as soon as 30-day LER NUREG

Event or Condition practical and in all cases practical and in all cases Sect.

within 1 hour within 4 hours

EMERGENCY CLASS Immediately after notification of Note-Although not specifically 3.1.1

State and local authorities, but mentioned in 10 CFR 50.73,

no later than 1 hour after many emergency class events

declaration of emergency class involve reportable situations

defined in licensee's emergency

plan [50.72(a)(1),(a)(2),(a)(3)

and (a)(4)]

TECHNICAL SPECIFICATIONS

(TS):

Plant shutdown (S/D) required by Initiation of S/D required by TS Completion of S/D required by 3.2.1

TS [50.72(b)(1)(I)(A)] TS [50.73(a)(2)(I)(A)]

Operation or condition 3.2.2

TS prohibited operations or prohibited by TS

conditions [50.73(a)(2)(I)(B)]

TS deviation authorized by Deviation from TS authorized by Criterion [50.73(a)(2)(l)(C)] 3.2.3

50.54(x) 50.54(x) [50.72(b)(1)(I)(B)] same as ENS 1 hour

DEGRADED CONDITION;

UNANALYZED CONDITION,

OUTSIDE DESIGN BASIS,

NOT COVERED BY

PROCEDURES:

Plant, including its principal safety During operation, serious Found while shut down; had it Either in operation or S/D, 3.2.4,

barriers, seriously degraded degradation of plant including its been found in operation, would condition of plant, including 3.3.1

principal safety barriers have been seriously degraded principal safety barriers,

[50.72(b)(1 [50.72(b)(2)(1)) seriously degraded

[50.73(a)(2)(ii)]
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Table 1 (continued)

ENS notification as soon as ENS notification as soon as NUREG

Event or Condition practical and in all cases practical and in all cases 30-day LER Sect.

within I hour within 4 hours

DEGRADED...(CONTINUED):

Plant in unanalyzed condition During operation, plant in Found while shut down; had it Either in operation or S/D, 3.2.4, significantly compromising plant unanalyzed condition, been found in operation, would unanalyzed condition 3.3.1

safety significantly compromising plant have been unanalyzed condition significantly compromising safety [50.72(b)(1)(ii)(A)] that significantly compromises plant safety [50.73(a)(2)(ii)(A)]

plant safety [50.72(b)(2)(1)]

Plant outside design basis of plant During operation, plant in While either in operation or 3.2.4,

condition outside design basis S/D, plant was in condition 3.3.1

[50.72(b)(1)(ii)(B)] outside design basis

[50.73(a)(2)(ii)(B)]

Plant in condition not covered by During operation, plant in While either in operation or 3.2.4, operating and emergency condition not covered by S/D, plant was in condition not 3.3.1

procedures operating and emergency covered by operating and

procedures [50.72(b)(1)(ii)(C)] emergency procedures

[50.73(a)(2)(ii)(C)]

EXTERNAL THREAT TO PLANT Any natural phenomenon or Criterion [50.73(a)(2)(iii)] 3.2.5

SAFETY other external condition that same as ENS 1 hour

poses an actual threat to the

safety of the plant or

significantly hampers site

personnel in performance of

duties necessary for its safe

operation [50.72(b)(1)(iii)]

EMERGENCY CORE COOLING A valid ECCS signal that results, Manual or automatic actuation Criterion [50.73(a)(2)(iv)] 3.2.6, SYSTEM (ECCS) DISCHARGE; or should have resulted, in of any ESF, including the encompasses both ENS 1 3.3.2

ACTUATION OF ANY ECCS discharge into the reactor reactor protection system hour and 4 hours

ENGINEERED SAFETY (ESF) coolant system [50.72(b)(1)(iv)] (RPS), occurs and was not

preplanned as part of a test or

reactor operation [50.72(b)(2)(ii)_
Table I (continued)

(0

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ENS notification as soon as ENS notification as soon as NUREG

Event or Condition practical and in all cases practical and in all cases 30-day LER Sect.

within 1 hour within 4 hours

EVENT THAT ALONE COULD Event or condition alone would Criterion [50.73(a)(2)(v)] same 3.3.3

HAVE PREVENTED have prevented fulfillment of as ENS 4 hours. Need not

FULFILLMENT OF A SAFETY safety function of system report individual component

FUNCTION needed for S/D of the reactor, failures under this paragraph

maintenance of a safe S/D if redundant equipment in

condition, residual heat removal same system was operable

(RHR), control of release of and available [50.73(a)(2)(vi)]

radioactive material, or

mitigation of the consequences

of an accident [50.72(b)(2)(iii)]

COMMON CAUSE OR Single cause or condition 3.3.4

CONDITION RESULTING IN caused inoperability of at

INDEPENDENT TRAINS OR least one independent train or

CHANNELS BECOMING channel in multiple systems or

INOPERABLE two independent trains and

channels in a single system

designed for safe S/D, RHR,

radiation release control, or

accident mitigation

[50.73(a)(2)(vii)]

RADIOACTIVE RELEASES:

Airborne radioactivity releases Airborne radioactivity released Criterion [50.73(a)(2)(viii)(A)] 3.3.5

to an unrestricted area exceeds same as ENS 4 hours.

20x the concentration specified

in 10 CFR 20, Appendix B,

Table 2, averaged over 1 hour

[50.72(b)(2)(iv)(A)]
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Table 1 (continued)

ENS notification as soon as ENS notification as soon as NUREG

Event or Condition practical and in all cases practical and in all cases 30-day LER Sect.

within I hour within 4 hours

RELEASES (CONTINUED):

Liquid effluent releases Liquid effluent released to an

unrestricted area exceeds 20x

the concentration specified in 10 Criterion [50.73(a)(2)(viii)(B)]

CFR 20, Appendix B, Table 2, same as ENS 4 hours. 3.3.5

for all radionuclides except

tritium and dissolved noble

gases, averaged over 1 hour

[50.72(b)(2)(iv)(B)].

INTERNAL THREAT TO PLANT Any event that poses an actual Criterion [50.73(a)(2)(x)] same 3.2.8

SAFETY threat to the safety of the plant as ENS 1 hour

or significantly hampers site

personnel in the conduct of safe

operation [50.72(b)(1)(vi)]

LOSS OF EMERGENCY A major loss of capability occurs 3.2.7

ASSESSMENT, OFFSITE for emergency assessment,

RESPONSE, OR offsite response, or

COMMUNICATIONS communications [50.72(b)(1)(v)]

CAPABILITY

TRANSPORT OF A radioactively contaminated 3.3.6

CONTAMINATED PERSON TO person is transported to an

OFFSITE MEDICAL FACILITY offsite medical facility

[50.72(b)(2)(v)]
Table I (continued)

ENS notification as soon as ENS notification as soon as NUREG

Event or Condition practical and in all cases practical and in all cases 30-day LER Sect

within 1 hour within 4 hours

NEWS RELEASE/OTHER A news release is planned or 3.3.7

GOVERNMENT NOTIFICATIONS other government agencies have

been or will be notified of an

event related to the health and

safety of the public or onsite

personnel, or the protection of

the environment [50.72(b)(2)(vi)

DEGRADED SPENT FUEL A defect in any spent fuel storage 3.3.8

STORAGE CASK OR cask structure, system, or

CONFINEMENT SYSTEM component that is important to

safety [50.72(b)(2)(vii)(A)J. A

significant reduction in the

effectiveness of any spent fuel

storage cask confinement system

during use of the storage cask

under a general licensee issued

under

10 CFR 72.210

[50.72(b)(2)(vii)(B)]

Followup Notification (Section 3.4):

After making a 1-hour or 4-hour notification, licensees are required to immediately notify the NRC Operations Center if any of the following occur

* plant conditions worsen [50.72(c)(1)(1)J, emergency classification changed [50.72(c)(1)(ii)], or emergency class terminated [50.72(cXl)(iii)];

0 the results of ensuing evaluations or assessments of plant conditions are obtained [50.72(cX2XI)];

0 the effectiveness of response or protective measures taken becomes known [50.72(c)(2Xii)];

* Information related to plant behavior is not understood [50.72(cX2Xiii)];

In addition, if requested by the NRC, maintain an open, continuous communication channel with the NRC Operations Center [50.72(cX3)].

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2 REPORTING AREAS WARRANTING SPECIAL MENTION

This section clarifies specific areas that are applicable to many reporting criteria or that

historically appear to be subject to varied interpretations.

2.1 Engineering Judgment

The reportability of many events and conditions is self evident. However, the reportability of

other events and conditions may not be readily apparent and the use of engineering judgment

is involved in determining reportability.

Engineering judgment may include either a documented engineering analysis or a judgment by

a technically qualified individual, depending on the complexity, seriousness, and nature of the

event or condition. A documented engineering analysis is not a requirement for all events or

conditions, but it would be appropriate for particularly complex situations. In addition, although

not required by the rule, it may be prudent to record in writing that a judgment was exercised by

identifying the individual making the judgment, the date made, and briefly documenting the

basis for this judgment. In any case, the staff considers that the use of engineering judgment

Impl0-ie a*I loicltho-ug*ht jprocess that supports the judgment.

2.2 Differences in Tense Between 10 CFR 50.72 and 50.73

The present tense was used in 10 CFR 50.72 because the event or condition generally would

be ongoing at the time of reporting. The past tense was used in 10 CFR 50.73 because the

event or condition normally would be past when an LER was written.

This difference creates some confusion over the reportability under 10 CFR 50.72 of events not

related to an ongoing event or discovered as the result of an event review. In other cases,

questions are raised regarding the need for a 10 CFR 50.73 report. Where the tense is

relevant to reportability, it is addressed in the specific criterion in Section 3 of this report.

2.3 Multiple Failures and Related Events

More than one failure or event may be reported in a single ENS notification or LER if (1) the

failures or events are related (have the same general cause or consequences) and (2) they

occurred during a single activity (e.g., test program) over a reasonably short time (within the

ENS reporting time limit for ENS reports, or within the first 30 days of discovery of the first

reportable event for LER reporting).

For an outage that lasts longer than 30 days, such as 60 days, similar events that are part of

the same activity or test program and are therefore related may be reported as a single LER.

13 NUREG-1022, Rev. 1
To the extent feasible, report failures that occurred within the first 30 days of discovery of the

first failure on one LER. State in the LER text that a supplement to the LER will be submitted

when the test is completed. Include all the failures, including those reported in the original LER,

in the revised LER (i.e., the revised LER should stand alone).

Generally, LERs are intended to address specific events and plant conditions. Thus, unrelated

events or conditions should not be reported in one LER. Also, an LER revision should not be

used to report subsequent failures of the same or like components that are the result of a

different cause or for separate events or activities.

Unrelated failures or events should be reported as separate ENS notifications to be given

unique ENS numbers by the NRC. However, multiple ENS notifications may be addressed in a

single telephone call.

2.4 Deficiencies Discovered During Engineering Reviews or Inspections

As indicated in NUREG-1397, "An Assessment of Design Control Practices and Design

Reconstitution Programs in the Nuclear Power Industry," February 1991, Section 4.3.2, the

reporting requirements specified in 10 CFR 50.9, 50.72, and 50.73 apply equally to

discrepancies discovered during design document reconstitution (DDR) programs, designbases documentation reviews (DBDRs), and other similar engineering reviews. There is no

basis for treating discrepancies discovered during such reviews differently from any other

reportable item.

Licensees should evaluate the reportability of suspected but unsubstantiated discrepancies

discovered during such a review program in the same manner as other potentially reportable

items. See Section 2.11 for discussion of reporting time limits and discovery dates.

2.5 Engineered Safety Features Actuations

Systems typically reported under this criterion include the. systems listed in Table 2. TheNRC

staff considers these systems to be a reasonable interpretation of what constitutes systems

"provided to mitigate the consequences of a significant event." Further guidance beyond that

provided in this revision is being deferred pending rulemaking to address issues such as

whether the rule should list specific systems or provide general guidance. See Section 3.3.2 for

further discussion of this matter.

2.6 Events Discussed with the NRC Staff

Some licensee personnel have erroneously believed that if a reportable event or condition had

been discussed with the resident inspector or other NRC staff, there was no need to report

under 10 CFR 50.72 and 50.73 because the NRC was aware of the situation. Some licensee

personnel have also expressed a similar understanding for cases in which the NRC staff

identified a reportable event or condition to the licensee via inspection or assessment activities.

Such means of reporting do not satisfy 10 CFR 50.72 and 50.73. The requirement is to report

to the ENS and LER systems events or conditions meeting the criteria stated in the rules.

NUREG-1022, Rev. 1 14
2.7 Multiple Component Failures

There have been cases in which licensees have not reported multiple, sequentially discovered

failures of systems or components occurring during planned testing. This situation was

identified as a generic concern on April 13, 1985, in NRC Information Notice (IN) 85-27,

"Notifications to the NRC Operations Center and Reporting Events in Licensee Event Reports,"

regarding the reportability of multiple events in accordance with §§50.72(b)(2)(iii) and

50.73(a)(2)(v) (event or condition that alone could prevent fulfillment of a safety function). (This

reporting criterion is discussed in Section 3.3.3 of this report.)

IN 85-27 described multiple failures of a reactor protection system during control rod insertion

testing of a reactor at power. One of the control rods stuck. Subsequent testing identified 3

additional rods that would not insert (scram) into the core and 11 control rods that had an initial

hesitation before insertion. The licensee considered each failure as a single random failure;

thus each was determined not to be reportable. Subsequent assessments indicated that the

instrument air system, which was to be oil-free, was contaminated with oil that was causing the

scram solenoid valves to fail. While the failure of a single rod to insert may not cause a

reasonable doubt that other rods would fail to insert, the failure of more than one rod does

cause a reasonable doubt that other rods could be affected, thus affecting the safety function of

the rods.

As indicated in IN 85-27, multiple failures of redundant components of a safety system are

sufficient reason to expect that the failure mechanism, even though not known, could prevent

the fulfillment of the safety function.

2.8 Preparation of Licensee Event Reports (LERs)

This revision includes nlew guidainceto address consistency .f information provided in LERs

Which is usedto.understa~nd*reported events. Some of this guidance was not specifically

addressed in the second draft that was published for comment. It is included to help ensure

that consistent information is provided regarding human factors analysis and risk-informed

regulatory programs such as accident sequence precursor (ASP) analysis and equipment

reliability estimates. It does not affect the decision as to whether or not an event must be

reported. See Section 5 for further discussion.

2.9 Voluntary Reporting

Information that does not meet the reporting criteria of 10 CFR 50.72 and 50.73 may be

reportable under other requirements such as 10 CFR 50.9, 20.2202, 20.2203, 50.36, 72.74,

72.216, 73.71, and Part 21. In particular, 10 CFR 50.9 (b) states "Each applicant or licensee

shall notify the Commission of information identified by the applicant or licensee as having for

the regulated activity a significant implication for public health and safety or common defense

and security." This applies to information which is not already required by other reporting or

updating requirements. Notification must be made to the Administrator of the appropriate

Regional Office within two working days of identifying the information. Reporting pursuant to

15 NUREG-1022, Rev. 1
§50.9 is required, not voluntary.") Voluntary reporting, as discussed in the following

paragraphs, pertains to information of lesser significance than described in §50.9(b).

The Statement of Considerations for 10 CFR 50.73 states "...licensees are permitted and

encouraged to report any event or condition that does not meet the criteria contained in

§50.73(a), if the licensee believes that the event or condition might be of safety significance or

of generic interest or concern. Reporting requirements aside, assurance of safe operation of all

plants depends on accurate and complete reporting by each licensee of all events having

potential safety significance."'2) Instructions for completing voluntary LERs are discussed in

Section 5.1.5 of this report. In addition, voluntary reporting is encouraged under 10 CFR 50.72,

as discussed in Section 4.2.2 of this report.

The NRC staff encourages voluntary LERs rather than information letters to report operational

events that do not meet the criteria contained in 10 CFR 50.73. The LER format is preferable

because it provides for the information needed to support NRC review of the event and

facilitates administrative processing, including data entry.

2.10 Retraction/Cancellation of Event Reports

Licensees have expressed concerns about the counting of event reports, both ENS notifications

and LERs. The NRC staff has indicated that its interest is in evaluating the reported

information, not in simply counting the number of events reported. While event reports may be

formally withdrawn, the staff has often found the information reported useful and has

maintained the information on file with the withdrawal notation.

If a licensee so chooses, an ENS notification can be retracted via a follow-up ENS call..LER

retractions should be made by letter. The retractions and cancellations are further discussed in

Section 4 for ENS notifications and Section 5 for LERs. Sound, logical bases for the withdrawal

should be communicated with the retraction or cancellation., (Example 3 in Section 3.3.1

illustrates a case where there were sound reasons for a retraction. The last event under

Example 1 in Section 3.3.2 illustrates a case where the reasons for retraction were not

adequate.)

2.11 Time Limits for Reporting

Reporting times in 10 CFR 50.72 are keyed to the occurrence of the event or condition, as

described below.

(') As indicated in the Statement of Considerations for §50.9, "A licensee cannot evade

the rule by never 'finding' information to be significant. The fact that a licensee considers

information to be significant can be established, for example, by the actions taken by the

licensee to evaluate that information." 59 FR 49362, December 31, 1987.

(2) 48 FR 33853, July 26,1983.

NUREG-1022, Rev. 1 16
Section 50.72(a)(3) requires ENS notification of the declaration of an Emergency Class

"...immediately after notification of the appropriate State or local agencies and not later

than one-hour after the time the licensee declares one of the Emergency Classes."

Section 50.72(b)(1) requires ENS notification for specific types of events and conditions

"...as soon as practical and in all cases, within one-hour of the occurrence of any of the

following: ......

Section 50.72(b)(2) requires ENS notification for specific types of events and conditions

"...as soon as practical and in all cases, within four hours of the occurrence of any of the

following: ......

Section 50.73 requires submittal of an LER "within 30 days after the discovery" of a reportable

event.

Many reportable events are discovered when they occur. However, if the event is discovered at

some later time, the discovery date is when the reportability clock starts under 10 CFR 50.73.

Discovery date is generally the date when the event was discovered rather than the date when

an evaluation of the event is completed. For example, if a technician sees a problem, but a

delay occurs before an engineer or supervisor has a chance to review the situation, the

discovery date (which starts the 30-day clock) is the date that the technician sees a problem.

In some cases, such as discovery of an existing but previously unrecognized condition, it may

be necessary to undertake an evaluation in order to determine if an event or condition is

reportable. If so, the guidance provided in Generic Letter 91-18, "Information to Licensees

Regarding two NRC Inspection Manual Sections on Resolution of Degraded and

Nonconforming Conditions and on Operability," which applies primarily to operability

determinations, is appropriate for reportability determinations as well. This guidance indicates

that, whenever reasonable expectation that the equipment in question is operable no longer

exists, or significant doubts begin to arise, appropriate actions, including reporting, should be

taken.

2.12 Outside Design Basis

This revision povides new guidance and several examples for this reporting criterion and two.

related criteria, seriously degraded condition and unanalyzed condition that seriously

compromised plant safety. Further guidance beyond that provided in this revision is being

deferred pending rulemaking to address issues such as: (1) one-hour reporting for design basis

issues, (2) significance testing for reporting design basis issues, and (3) scope of plant design

basis. See Section 3.2.4 of this report for further discussion.

17 NUREG-1022, Rev. 1

3 SPECIFIC REPORTING GUIDELINES

This section addresses the specific requirements of each part of the rules cited for immediate

notification of an event under 10 CFR 50.72 via the ENS and 30-day written reports under 10

CFR 50.73 via LERs. The section is divided into four parts. Section 3.1 gives the general

requirements for reporting, Section 3.2 gives the criteria for 1-hour notifications and 30-day

reports, Section 3.3 gives the criteria for 4-hour notifications and 30-day reports, and Section

3.4 addresses followup notifications.

The sequential scheme of 10 CFR 50.72 is used, which generally categorizes the times for

reporting by the relative importance of the event or condition. Because considerable overlap

exists between the various reporting criteria in each rule, the associated requirements for

licensee event reporting (10 CFR 50.73) are given coincidentally. Differences in the wording of

the comparable parts of the rules are underlined. In several instances, the wording of the two

rules is the same except for verb tense. A discussion of reporting guidelines and examples

follow each citation of specific parts of the rules. Brief examples occasionally are given in the

discussion for clarification; however, expanded examples for each part of the rules are

discussed under "Examples." The descriptions in the expanded examples have been taken

from actual operational experience and have been condensed to illustrate specific aspects of

reportability.

The reporting requirements in each of the two rules are not mutually exclusive, and many

events and conditions are reportable under more than one criterion. Therefore, it is important

to first recognize whether an event or condition is reportable under at least one criterion, and

then to identify other applicable criteria. When the report is made to the NRC, applicable

criteria should be cited.

19 NUREG-1022, Rev. 1
3.1 Section 50.72 and 50.73 General Requirements

3.1.1 Immediate Notification Requirements

§50.72(a) General Requirements1

"(1) Each nuclear power reactor licensee licensed under

§50.21(b) or §50.22 of this part shall notify the NRC

Operations Center via the Emergency Notification System of:

(i) The declaration of any of the Emergency Classes

specified in the licensee's approved Emergency Plan;2 or

(ii) Of those non-Emergency events specified in paragraph

(b) of this section.

(2) If the Emergency Notification System is inoperative, the

licensee shall make the required notifications via commercial

telephone service, other dedicated telephone system, or any

other method which will ensure that a report is made as soon

as practical to the NRC Operations Center.3

(3) The licensee shall notify the NRC immediately after

notification of the appropriate State or local agencies and not

later than one hour after the time the licensee declares one of

the Emergency Classes.

(4) The licensee shall activate the Emergency Response

Data System (ERDS)5 as soon as possible but not later than

one hour after declaring an emergency class of alert, site area

emergency, or general emergency. The ERDS may also be

activated by the licensee during emergency drills or exercises

if the licensee's computer system has capability to transmit

the exercise data."

"1 Other requirements for immediate notification of the NRC

by licensed operating nuclear power reactors are contained

elsewhere in this chapter, in particular, §§ 20.1906, 20.2202,

50.36, and 73.71.

2 These Emergency Classes are addressed in Appendix E

of this part.

3 Commercial telephone number of the NRC Operations

Center is (301) 816-5100."

4 [Reserved]

5 Requirements for ERDS are addressed in Appendix E,

Section VI."

(Continued on next page)

10 CFR 50.73

[If the event or condition

that was the basis for

the Emergency Class

declaration met one or

more of the 10 CFR

50.73 reporting criteria,

an LER is required.]

NUREG-1022, Rev. 1 20
50.72(a) (Continued)

"(5) When making a report under paragraph (a)(3) of this

section, the licensee shall identify

(i) The Emergency Class declared; or

(ii) Either paragraph (b)(1), "One-Hour Report," or

paragraph (b)(2), "Four-Hour Report," as the paragraph of this

section requiring notification of the Non-Emergency Event."

Discussion

Appendix E to 10 CFR Part 50, Section IV (C), "Activation of Emergency Organization,"

establishes four emergency classes for nuclear power plants: Notification of Unusual Event,

Alert, Site Area Emergency, and General Emergency. NUREG-0654/FEMA-REP-1, Revision 1,

"Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and

Preparedness in Support of Nuclear Power Plants" (March 1987), and more recently,

NUMARCINESP-007, Revision 2, "Methodology for Development of Emergency Action Levels"

(January 1992), provides the basis for these emergency classes and numerous examples of the

events and conditions typical of each emergency class. Licensees use this guidance in

preparing their emergency plans. Use of these four emergency class terms in the ENS

notification will help the NRC recognize the significance of an emergency. Time frames

specified for notification in §50.72(a) use the words "immediately" and "not later than one hour"

to ensure the Commission can fulfill its responsibilities during and following the most serious

events.

Occasionally, a licensee may discover that an event or condiftion had exilsted Which met the

emergency plan criteria but that no emergency had been declared and the basis for the

emergency class no longer exists at the time of this discovery. This may be due to a rapidly

concluded event or an oversight in the emergency made madeduring the event or it

may be determined during a post-event review. Frequently, in cases of this nature, which were

discovered after the fact, licensees have declared the emergency class, immediately terminated

the emergency class and then made the appropriate notifications. How~ever, the staff does not

consider actual declaration of the emergency class to be necessary in these circumstances; an

ENS notification (or an ENS update if the event was previo.us~ly reported but misclassified)

within one hour of the discovery of the und~eclared (or misclassified) event will provide an

acceptable altemnative.0)

(3 Notification of the State and local emergency response organizations should be

made in accordance with the arrangements made between the licensee and offsite

organizations.

21 NUREG-1022, Rev. I
3.1.2 Licensee Event Report System

10 CFR 50.72

[Bases for ENS notifications

(e.g., regardless of plant

status), are the same as 10

CFR 50.73 where the two

rules are complementary.]

§50.73(a)(1)

"The holder of an operating license for a nuclear power

plant (licensee) shall submit a Licensee Event Report

(LER) for any event of the type described in this paragraph

within 30 days after the discovery of the event. Unless

otherwise specified in this section, the licensee shall report

an event regardless of the plant mode or power level, and

regardless of the significance of the structure, system, or

component that initiated the event."

Discussion

Unless otherwise specified, this part of the rule requires reporting of an event regardless of the

plant mode or power level and regardless of the significance of the structure, system, or

component that initiated the event. These considerations also are implicit in 10 CFR 50.72

where the two rules are complementary.

NUREG-1022, Rev. 1 22
3.2 One-hour ENS Notifications and 30-Day LERs

This section addresses §50.72(b)(1) 1-hour notifications for non-emergency events and the

associated 10 CFR 50.73 written reports. If not reported as a declaration of an emergency class

under §50.72(a), licensees are to notify the NRC as soon as practical and in all cases within I

hour of the occurrence of any of the events specified in §50.72(b)(1) and to submit an LER, if

specified.

In addition to similar reporting criteria under both 10 CFR 50.72 and 50.73, several

requirements for only 50.72 notifications or only LERs are included in this section because of

the sequential numbering scheme used. For example, operation or a condition prohibited by

the plant's technical specifications (TS), as discussed in Section 3.2.2, requires an LER but no

ENS notification, while loss of emergency assessment, response or communications capability,

as discussed in Section 3.2.7, requires an ENS notification but no LER.

23 NUREG-1022, Rev. 1
3.2.1 Plant Shutdown Required by Technical Specifications

§50.72(b)(1)(i)(A) §50.73(a)(2)(i)(A)

Licensees shall rerq_: "The initiation of any Licensees shall submit a Licensee Event

nuclear plant shutdown required by the Report on: "The completion of any nuclear

plant's Technical Specifications." plant shutdown required by the plant's

Technical Specifications."

If not reported as an emergency under §50.72(a), licensees are required to report the initiation

of a plant shutdown required by TS to the NRC via the ENS as soon as practical and in all

cases within 1 hour of the initiation of a plant shutdown required by TS to the NRC via the ENS.

If the shutdown is completed, licensees are required to submit an LER within 30 days.

Discussion

The §50.72 reporting requirement is intended to capture those events for which TS require the

initiation of reactor shutdown to provide the NRC with early warning of safety significant

conditions serious enough to warrant that the plant be shut down.

For §50.72 reporting purposes, the phrase "initiation of any nuclear plant shutdown" includes

action to start reducing reactor power, i.e., adding negative reactivity to achieve a nuclear plant

shutdown required by TS. The "initiation of any nuclear plant shutdown" does not include mode

changes required by TS if initiated after the plant is already in a shutdown conditionA reduction in power for some other purpose, not constituting initiation of a shutdown required

by TS, is not reportable under this criterion.

For §50.73 reporting purposes, the phrase "completion of any nuclear plant shutdown" is

defined as the point in time during a TS required shutdown when the plant enters the first

shutdown condition required by a limiting condition for operations (LCO) e.g., hot standby

{Mode 3] for PWRs with the standard technical specifications (STS). For example, if at.0200

hours a plant enters an LCO action statement that states, "restore the inoperable channel to

operable status within 12 hours or be in at least Hot Standby within the next 6 hours," the plant

must be shut down (i.e., at least in hot standby) by 2000 hours. An LER is required if the

inoperable channel is not returned to operable status by 2000 hours and the plant enters hot

standby.

An LER is not required if a failure was or could have been corrected before a plant has

completed shutdown (as discussed above) and no other criteria in §50.73 apply.

NUREG-1022, Rev. 1 24
Examples

(1) Initiation of a TS-Required Plant Shutdown

While operating at 100-percent power, one of the battery chargers, which feeds a 125

Vdc vital bus, failed during a surveillance test. The battery charger was declared

inoperable, placing the plant in a 2-hour LCO to return the battery charger to an

operable status or commence a TS-required plant shutdown. Licensee personnel

started reducing reactor power to achieve a nuclear plant shutdown required by a TS

when they were unable to complete repairs to the inoperable battery charger in the 2

hours allowed. The cause of the battery charger failure was subsequently identified and

repaired. Upon completion of surveillance testing, the battery charger was returned to

service and the TS required plant shutdown was stopped at 96-percent power.

The licensee made an ENS notification because of the initiation of a TS-required plant

shutdown An LER was not submitted under this criterion since the failed battery charger

was corrected before the plant completed shutdown.

(2) Initiation and Completion of a TS-Required Plant Shutdown

During startup of a PWR plant with reactor power in the intermediate range, two of the

four reactor coolant pumps (RCPs) tripped when the station power transformer

supplying power, deenergized. With less than four RCPs operating, the plant entered a

1-hour LCO to be in hot standby. Control rods were manually inserted to place the plant

in a shutdown condition.

The licensee made an ENS notification because of the initiation of a TS-required plant

shutdown. An LER was submitted within 30 days because of the completion of the TSrequired plant shutdown.

(3) Failure that was or could have been corrected before a plant has completed shut down.

S Question:

What about the situation where you have seven days to fix a component or be

shut down, but the plant must be shut down to fix the component? Assume the

plant shuts down, the component is fixed, and the plant returns to power prior to

the end of the seven day period. Is that situation reportable?

Answer:

No. If the shutdown was not required by the Technical Specifications, it need not

be reported. However, other criteria in 50.73 may apply and may require that the

event be reported.

25 NUREG-1022, Rev. 1
0 Question:

Suppose that there are seven days to fix a problem and it is likely the problem

can be fixed during this time period. However, the plant management elects to

shut down and fix this problem and other problems. It an LER required?

Answer:

Some judgment is required. An LER is not required if the situation could have

been corrected before the plant was required to be shut down, and no other

criteria in 50.73 apply. The shut down is reportable, however, if the situation

could not have been corrected before the plant was required to be shut down, or

if other criteria of 50.73 apply.

NUREG-1022, Rev. 1 26
3.2.2 Technical Specification Prohibited Operation or Condition

10 CFR 50.72 §50.73(a)(2)(i)(B)

[There is no corresponding Part 50.72 requirement. Licensees shall report: "any

However, for certain operations or conditions operation or condition prohibited

prohibited by a plant's TS, other reporting by the plant's Technical

requirements may apply, such as 50.72(b)(1)(ii) and Specifications."

(b)(2)(iii); 50.36(c)(1) and (2); 20.2202; and 20.2203.]

Licensees are required to submit an LER within 30 days for any operation or condition

prohibited by technical specifications.

Discussion

Section 50.73(a)(2)(i)(B) requires any operation or condition that is prohibited by the plant's TS

to be reported in an LER. The five specific TS categories defined in 10 CFR 50.36(c),

"Technical Specifications," are discussed below. In addition, based on past experience,

guidelines are provided for reporting entry into TS 3.0.3 [ISTS (4) Limiting Condition for

Operation (LCO) 3.0.3]; missed or deficient tests required by the American Society of

Mechanical Engineers (ASME) Section XI, Inservice Testing (IST) and Inservice Inspection

(ISI), and by STS 4.0.5, or equivalent; and other operations or conditions prohibited by TS, such

as fire protection.

The LER rule does not address violations of license conditions contained in documents other

than the TS. Such notifications are reportable as specified in a plant's license or other

applicable documents.

(1) Safety Limits and Limiting Safety System Settings

Section 50.36(c)(1) outlines the reporting requirements in TS when nuclear reactor

safety limits or limiting safety system settings are exceeded and identifies that such

reports are to be made under 50.72 and 50.73.

(4) To be consistent with the improved Standard Technical Specifications (ISTS)

discussed in the NUREG-1430 through NUREG-1434 (e.g., NUREG-1431, Vol. 1, Standard

Technical Specifications - Westinghouse Plants, September 1992) references to appropriate

sections in these ISTS have been included throughout this section of NUREG-1022. The

designation used here for references to such sections is "ISTS" followed by the appropriate

section number.

27 NUREG-1022, Rev. 1
(2) Limiting Conditions for Operation

Section 50.36(c)(2) outlines LCOs in TS. Certain TS contain LCO statements that

include action statements (required actions and associated completion time in ISTS) to

provide constraints on the length of time components or systems may remain inoperable

or out of service before the plant must shut down or other compensatory measures must

be taken. Such time constraints are based on the safety significance of the component

or system being removed from service.

An LER is required if the conditions of an LCO are not met, e.g., by exceeding action

statement constraints (not meeting required actions and associated completion times in

ISTS).

The LCO allows a plant a specific time interval referred to as the allowed outage time (or

completion time in ISTS) to accomplish corrective actions (e.g., restoration of

equipment, testing of other equipment, and/or an orderly shutdown to either the hot- or

cold-shutdown mode or operating condition).

If a condition existed for a time longer than permitted by the TS [i.e., greater than the

allowed outage time (or completion time in ISTS)] it must be reported even if the

condition was not discovered until after the allowable time had elapsed and the condition

was rectified immediately upon discovery. This guidance is consistent with that

previously given. (For the purpose of this discussion, it is assumed that there was firm

evidence that a condition prohibited by TS existed before discovery, for a time longer

than permitted by TS.)

(3) TS Surveillance Requirements

Section 50.36(c)(3) outlines surveillance requirements in TS which assure (1) necessary

quality of systems and components, (2) operation within safety limits, and (3) that the

limiting conditions for operation will be met. For the purpose of evaluating the

reportability of discrepancies found during TS surveillance tests, an operation or

condition prohibited by the TS existed and is reportable if the time of equipment

inoperability exceeded the LCO allowed outage time (or completion time for restoration

of equipment in ISTS). It should be assumed that the discrepancy occurred at the time

of its discovery unless there is firm evidence, based on a review of relevant information

(e.g., the'equipment history and cause of failure):to believe that the discrepancy existed

previously. As discussed in Example 5, evaluation of multiple similar failures may

indicate that a condition has persisted for some time.

Missed surveillance tests are reportable when the surveillance interval plus allowed

surveillance interval extension, e.g., STS section 4.0.2 (or ISTS SR 3.0.2), plus the LCO

action statement time is exceeded.4) 5 This means that a condition prohibited by TS

{s' The Statement of Considerations for the final rule (48 FR 33855, July 28, 1983,

Second column) states, in part, ".. . if a condition that was prohibited by the Technical

NUREG-1022, Rev. 1 28
existed for a period of time longer than allowed by TS. The event is reportable even

though the surveillance is subsequently satisfactorily performed.06

Some .plants have TS which alltow a delay of up to 24 hours In declaring antLCO or TS

requirements not met if it is found that a surveillance was not performed within its

specified frequency or interval. However, failure to perform a surveillance within its

frequency or interval is still reportable. The additional delay in declaring the -LCO not

met does not change the fact that the condition existed longer than allowed by TS. The

delay merely specifies appropriate remedial action.t

(4) Design Features

Section 50.36(c)(4) indicates that design features to be included in TS are those

features of the facility such as materials of construction or geometric arrangements

which, if altered or modified, would have a significant effect on safety and are not

covered by items (1) through (3) above.

Reportability requirements related to design features are included in other sections of 10

CFR 50.72 and 50.73.

(5) TS Administrative Requirements, Including Radiological Controls, Required by Section 6

of the STS, or Equivalent

Section 6 of the STS (Section 5 of ISTS), or its equivalent, has a number of

administrative requirements such as organizational structure, the required number of

personnel on shift, the maximum hours of work permitted during a specific interval of

time, and the requirement to have, maintain, and implement certain specified

procedures.

Failuýr~et3 mzc6,t 3ueh odministrti~ eg6U~irýmneb la priothibitedby th&ýM mS. hcthcrit

Ipeifiati as an LeR dapernds upe lon whgeitr theault permed in a Tehnicthe by-the LER

ful .if the vitlmtien of the onditionws no dScredsultsin TdfevnSifthe apefatiern

allw betiehd bylapstheand th ioepteiybleViolationsof an administrative TS in and of itself does not necessarily constitute a

reportable condition ("operation or condition prohibited by the plant's TS"). This

Specifications existed for a period of time longer than that permitted by the Technical

Specifications, it must be reported even if the condition was not discovered until after the

allowable time had elapsed and the condition was rectified immediately after discovery."

{B) This guidance is only intended to define when the matter becomes reportable under

this specific reporting criterion ("operation or condition prohibited by the plant's TS"). It is not

intended to define when a TS violation occurs, when a system must actually be declared

inoperable, when the surveillance must be completed, or when the plant must be shutdown.

These matters are discussed further in GL 87-09, GL 91-18, TS 4.0.2, and ISTS SR 3.0.3.

29 NUREG-1022, Rev. I
reporting requirement deals with matters affecting plant operation more substantially

and more directly than matters that are mainly administrative. Failure to meet

administrative TS requirements is reportable only if it results in violations of equipment

operability requirements, or had a similar; detrimental effect on a licensees ability to

safely operate the plant.

For example, operation with less than the required number of people on shift would

constitute operation prohibited by the TS. However, a change in the plant's

organizational structure that has not yet been approved as a Technical Specification

change would not.

An administrative procedure violation or failure to implement a procedure, such as

failure to lock a high radiation area door, that does not have a direct impact on the safe

operation of the plant, is generally not reportable under this criterion.

Radiological conditions and events that are reportable are defined in 10 CFR 20.2202

and 20.2203. Redundant reporting is not required.

(6) Entry into STS 3.0.3

STS 3.0.3 (ISTS LCO 3.0.3), or its equivalent, establishes requirements for actions

when an LCO is not met and no action statement is provided. Entry into STS 3.0.3 is

considered to be the action taken, as required, when operations or conditions required

by TS LCO action statements (ISTS required actions and associated completion times)

are not met. Thus, entry into STS 3.0.3 (ISTS LCO 3.0.3) for any reason or justification

is reportable.

(7) Missed Tests Required by ASME Section XI IST and ISI and by STS 4.0.5, or

Equivalent

Sections 50.55a(g) and 50.55a(f) require the implementation of ISI and IST programs in

accordance with the applicable edition of the ASME Code for those pumps and valves

whose function is required for safety. STS Section 4.0.5 (or an equivalent) covers these

testing requirements. (Generally, there is no comparable ISTS section.) Missed

ISTIISIlASME tests are reportable when the test interval plus any allowable extension

plus the LCO action time has been exceeded.

• The proposed rule would have required reporting when "a TS action statement is

not met." The wording of the final rule requires reporting "Any operation or condition prohibited

by the plant's Technical Specifications." The Statement of Considerations for the final rule

indicates that this change was made to accommodate plants that did not have requirements

specifically defined as action statements (48 FR 33855, July 26, 1983).

NUREG-1022, Rev. 1 30
(8) Fire Protection Systems When Required by TS

When operability requirements for fire protection systems are specified in TS they are

within the scope of this reporting criterion.

Examples

(1) LCO Exceeded

A licensee found a standby component with a 7-day LCO allowed outage time and

associated 8-hour shutdown action statement to be inoperable during a 30-day

surveillance test. (This is equivalent to a 7-day restoration completion time and an 8-

hour action completion time in ISTS.) Subsequent review indicated that the component

was assembled improperly during maintenance conducted 30 days previously and the

post-maintenance test was not adequate to identify the error. Thus, there was firm

evidence that the standby component had been inoperable for the entire 30 days.

An LER was required because the 7-day LCO allowed outage time and the shutdown

action statement time of 8 hours were exceeded. Had the inoperability been identified

and corrected within the 7-day LCO allowed outage time plus the 8-hour shutdown

action statement, the event would not be reportable.

(2) Missed Surveillance Tests

A licensee, with the plant in Mode 5 following a 10-month refueling outage, determined

that certain monthly TS surveillance tests, which were required to be performed

regardless of plant mode, had not been performed as required during the outage. The

STS 4.0.2 (equivalent to ISTS SR 3.0.2) extension was also exceeded. The

surveillance tests were immediately performed. An LER is required because the time

interval, including extensions permitted by TS, exceeded the TS surveillance interval

plus the LCO action statement times (equivalent to ISTS completion times).

(3) Entering STS 3.0.3

With essential water chillers (A) and (B) out of service, the only remaining operable

chiller (A/B) tripped. This condition caused the plant to enter STS 3.0.3 (equivalent to

ISTS LCO 3.0.3) for 1 hour until chiller (A) was restored to service and the temperature

was restored to within TS limits. An LER is required for this event because STS 3.0.3

was entered.

(4) Missed Tests Required by ASME Section XI IST and ISI, and by STS 4.0.5, or

Equivalent

Examples of potentially reportable conditions are failures to perform required activities

within specified times for those components governed by TS. Such activities include

stroke testing valves, testing valves in the position required for the performance of their

safety function, verifying motor-operated valve stroke times for both (open and closed)

31 NUREG-1022, Rev. 1
directions, using the proper test pressures to properly classify and test active valves and

to increase test frequency subsequent to obtaining test results that were below certain

threshold values. A missed testis reportable when the test interval plus any allowable

extension plus the LCO action time is exceeded.

(5) Multiple Test Failures

An example of multiple test failures involves the sequential testing of safety valves.

Sometimes multiple valves are found to lift with setpoints outside of TS limits.

As discussed above, discrepancies found in TS surveillance tests should be assumed to

occur at the time of the test unless there is firm evidence, based on a review of relevant

information (e.g., the equipment history and the cause of failure) to believe that the

discrepancy occurred earlier. However, the existence of similar discrepancies in

multiple valves is an indication that the discrepancies arose over a period of time.

Therefore, the condition existed during plant operation and the event is reportable under

§50.73(a)(2)(i)(B) "Any operation or condition prohibited by the plant's Technical

Specifications."

If the discrepancies are large enough that multiple valves are inoperable the event may

also be reportable under §50.73(a)(2)(vii) "Any event where a single cause or condition

caused at least one independent train or channel to become inoperable in multiple

systems or two independent trains or channels to become inoperable in a single

system ..... "

NUREG-1022, Rev. 1 32
3.2.3 Technical Specification Deviation per §50.54(x)

§50.72(b)(1 )(i)(B) §50.73(a)(2)(i)(C)

Licensees shall report: "Any deviation from Licensees shall report: "Any deviation from

the plant's Technical Specifications the plant's Technical Specifications

authorized pursuant to §50.54(x) of this authorized pursuant to §50.54(x) of this

part." part."

If not reported as an emergency under §50.72(a), licensees are required to report any such

deviation to the NRC via the ENS as soon as practical and in all cases within 1 hour. Licensees

are required to submit an LER within 30 days.

Discussion

10 CFR 50.54(x) generally permits licensees to take reasonable action in an emergency even

though the action departs from the license conditions or plant technical specifications if (1) the

action is immediately needed to protect the public health and safety, including plant personnel,

and (2) no action consistent with the license conditions and technical specifications is

immediately apparent that can provide adequate or equivalent protection. Deviations

authorized pursuant to 10 CFR 50.54(x) are reportable under this criterion.

Exam"le

With the plant at 100-percent power, the upper containment airlock inner door was opened to

allow a technician to exit from the containment while the upper airlock outer door was

inoperable, resulting in the loss of containment integrity. The upper airlock door was inoperable

pending retests following seal replacement. The technician was inside containment when the

lower airlock failed, requiring the technician to exit through the upper door.

The licensee decided to exercise the option allowed for under 10 CFR 50.54(x) and open the

upper containment airlock inner door. In this instance, immediate action was considered

necessary to protect the safety of the technician. The upper airlock was not scheduled to be

returned to operability for another 20 hours and the time to repair the lower airlock door was

unknown. When the action was completed the control room operators notified the NRC

Operations Center, in accordance with the reporting requirements of 10 CFR 50.72, that they

had exercised 10 CFR 50.54(x).

Subsequently, an LER was submitted in accordance with 10 CFR 50.73(a)(2)(i) {use of 10 CFR

50.54(x)} as well as 10 CFR 50.73(a)(2)(v) (event or condition that alone could prevent .... }.

33 NUREG-1022, Rev. 1
3.2.4 Operating Plant Found in Degraded or Unanalyzed Condition

§50.72(b)(1)(ii)

Licensees shall report: "Any event or

condition during operation that results in the

condition of the nuclear power plant,

including its principal safety barriers, being

seriously degraded; or results in the nuclear

power plant being:

(A) In an unanalyzed condition that

significantly compromises plant safety;

(B) In a condition that ij outside the design

basis of the plant; or

(C) In a condition not covered by the

plant's operating and emergency

procedures."

§50.73(a)(2)(ii)

Licensees shall report: "Any event or

condition that resulted in the condition of the

nuclear power plant, including its principal

safety barriers, being seriously degraded; or

that resulted in the nuclear power plant

being:

(A) In an unanalyzed condition that

significantly compromised plant safety;

(B) In a condition that was outside the

design basis of the plant; or

(C) In a condition not covered by the

plant's operating and emergency

procedures."

If not reported as an emergency under §50.72(a), licensees are required to report operation

under such a condition to the NRC via the ENS as soon as practical and in all cases within 1

hour. Licensees are required to submit an LER within 30 days.

Discussion

Reporting at the component, system, and structure level is required under 10 CFR

50.72(b)(1)(ii) and 50.73(a)(2)(ii) if the event or condition resulted in the plant being seriously

degraded, in an unanalyzed condition that significantly compromises plant safety, outside the

plant design basis, or in a condition not covered by the plant's procedures, as described in the

rule.

The discussions below provide further guidance on reportability under these criteria.

L( The condition of the nuclear power plant, including its principal safety barriers, being

seriously degraded.

As indicated in the Statements of Considerations, this paragraph includes material (e.g.,

metallurgical or chemical) problems that cause abnormal degradation of the principal

safety barriers (i.e., the fuel cladding, reactor coolant system pressure boundary, or the

containment). Examples of this type of situation include:

(a) Fuel cladding failures in the reactor, or in the storage pool, that exceed expected

values, or that are unique or widespread, or that are caused by unexpected

factors, and would involve a release of significant quantities of fission products.

NUREG-1022, Rev. 1 34
(b) Cracks and breaks in the piping or reactor vessel (steel or prestressed concrete)

or major components in the primary coolant circuit that have safety relevance

(steam generators, reactor coolant pumps, valves, etc).

(c) Significant welding or material defects in the primary coolant system.

(d) Serious temperature or pressure transients.

(e) Loss of relief and/or safety valve functions during operation.

(f) Loss of containment function or integrity including:

(i) Containment leakage rates exceeding the authorized limits.

(ii) Loss of containment isolation valve function during tests or operation.

(iii) Loss of main steam isolation valve function during test or operation, or

(iv) Loss of containment cooling capability.

Examples of events that the staff would consider reportable as significant reactor

coolant system welding or material defects include items which cannot be found

acceptable under ASMVE Section Xl, IWB-3600, "Analytical Evaluation of Flaws" or

ASME Section X, Table IeyB-3410-1, "Acceptance Standards."

Examples of events that the staff would consider reportable-as serious temperature Or

pressure transients include low temperature over pressure transients where the

pressure-temperature relationship violates Mpressure-emperature limits derived from

Appendixa Gto 10 CFR Part 50 (e.g., TS pressure-temperature curves).

Examples of events the staff would consider reportable as containment leakage t

exceeding authorized limits include containment leak rate tests where the total

containment -as-found, minimum-pathway leak rate exceeds the LCO in the facility's

SThe LCO typically employs La, which is defined in Appendix J to 10 CFR Part 50 as

the maximum allowable containment leak rate at pressure Pa, the calculated peak containment

internal pressure related to the design basis accident. Minimum-pathway leak rate means the

minimum leak rate that can be attributed to a penetration leakage path; for example, the smaller

of either the inboard or outboard valve's individual leak rates.

(9) For such a condition, an LER is generally required under 10 CFR 50.73(a)(2)(ii). If

the condition existed during operation, an ENS notification would also be required by

§50.72(b)(1)(ii) if found during operation or by §50.72(b)(2)(i) if found while shutdown.

35 NUREG-1022, Rev. I
L21 The nuclear power plant being in an unanalyzed condition that significantly

compromises plant safety.

As indicated in the Statements of Consideration:

"The Commission recognizes that the licensee may use engineering judgment

and experience to determine whether an unanalyzed condition existed. It is not

intended that this paragraph apply to minor variations in individual parameters, or

to problems concerning single pieces of equipment. For example, at any time,

one or more safety-related components may be out of service due to testing,

maintenance, or a fault that has not yet been repaired. Any trivial single failure

or minor error in performing surveillance tests could produce a situation in which

two or more often unrelated, safety-grade components are out-of-service.

Technically, this is an unanalyzed condition. However, these events should be

reported only if they involve functionally related components or if they

significantly compromise plant safety."t 10 )

'When applying engineering judgment, and there is a doubt regarding whether to

report or not, the Commission's policy is that licensees should make the report.,,O"1

"For example, small voids in systems designed to remove heat from the reactor

core which have been previously shown through analysis not to be safety

significant need not be reported. However, the accumulation of voids that could

inhibit the ability to adequately remove heat from the reactor core, particularly

under natural circulation conditions, would constitute an unanalyzed condition

and would be reportable."121

"In addition, voiding in instrument lines that results in an erroneous indication

causing the operator to misunderstand the true condition of the plant is also an

unanalyzed condition and should be reported."' 13

'

(3 The nuclear power plant being in a condition that is outside the design basis of the plant.

As indicated in 10 CFR 50.2, "Design bases means that information which identifies the

specific functions to be performed by a structure, system, or component of a facility, and

the specific values or ranges of values chosen for controlling parameters as reference

bounds for design. These values may be (1) restraints derived from generally accepted 'state of the art' practices for achieving functional goals, or (2) requirements derived

{1o) 48 FR 39042, August 29, 1983 and 48 FR 33856, July 26, 1983.

{' ~48 FR 39042, August 29, 1983.

(12) 48 FR 39042, August 29, 1983 and 48 FR 33856, July 26, 1983.

(13) 48 FR 39042, August 29, 1983 and 48 FR 33856, July 26, 1983.

NUREG-1022, Rev. 1 36
from analysis (based on calculation and/or experiments) of the effects of a postulated

accident for which a structure, system, or component must meet its functional goals."

(Emphasis added.)

Examples of events or conditions the staff considers reportable include errors in the

actual design, such as discovery that an ECCS design does not meet the single failure

criterion. They also include hardware problems such as discovery that high energy line

break restraints are not installed.

With regard to ECCS calculations, detailed reporting criteria are given in 10 CFR

50.46(a). That rule provides that a change or error correction that results in calculated

ECCS performance that does not meet the acceptance criteria {peak cladding

temperature, cladding oxidation, etc.) is a reportable event as described in 10 CFR

50.55(e), 50.72 and 50.73. Lesser changes or error corrections are the subject of other,

separate reports.

Violation of fire protection commitments regarding safe shutdown capabilty May indicate

that the plant is outside of its design basis, For example, if fire barriers are found to be

missing, osuch that the required degree of separation for redundant safe shutdown trains

is lacking, the plant Would be outside of the design ba~sis with respect to A..ppend.ix R to

10 CFR Part 50, On the other hand, ifta fire wrap, to Which, the licensee has committed,

is missing from- a safe shutdown train but another safe shutdown train is available in a

differe nt fire area, protecctd such that the required sepaatiogn fo sae shptedco n trains

is still provided, the plant would rot be outside ofitsdesi basis with respect to

Appendix RY')4

Another examnple of an event or condition that the staiff considers reportable is discovery that one train of a required two train safety system ha..s been incapal fpromn t

design function (intended safety finction) for an extended period of time during

,Operation. For example, in a two-train ECOS system, one train might be found with a

design flaw or with a component that would neve~r have- functiloned be~cau~seitw~as

installed incorrectly and a test that would reveal the problem was not performed, such

that the train was incapable of performing its design function. This would be consid~ered4

outside the design -basis because the system did not have suitable redundancpy.0 5)

c 1i aThe design basis with respect to Appendix R for protection of safe shutdown

capability is essentially the same as the required protection features. This is discussed in the

Statement of Considerations for Appendix R, Federal Register, November 19, 1980 (45 FR

76606). In particular, it is stated that "Because it is not possible to predict the specific

conditions under which fires may occur and propagate, the design basis protective features are

specified rather than the design basis fire."

(15) A minimum design feature is suitable redundancy meeting the single-failure

criterion as indicated in: (1) 10 CFR Part 50, Appendix A, Introduction and 10 CF5R 50,

Appendix A, Criterion 35; (2) 10 CFR 50, Appendix K, Item 1. D. 1; AND (3) FSAR commitments.

37 NUREG-1022, Rev. 1
It should be noted that these discussions concern events or conditions that actually

place the plant outside its design basis. They are not intended to. capture minor

problems such as: (1) cases of administrative inoperability, where a component is

declared inoperable because a surveillance test is overdue but the equipment is actually

capable of performing its design function, or (2) cases where the LCO allowed outage

time is exceeded by a modest amount(e~g., less than 25 percent). Such conditions

may, however, be reportable as conditions prohibited by the Technical Specifications, 10

CFR 50.73(a)(2)(i)(B),

(4) The nuclear power plant being in a condition not covered by the plant's operating and

emergency procedures.

This criterion points to events where the plant is in a condition outside the coverage of

its operating and emergency procedures. A straightforward example of this type of

event was the accident at Three Mile Island,

Examples

(1) Design Problem (ECCS Single Failure Vulnerability)

A minimum design feature for ECCS is suitable redundancy meeting the single-failure

criterion. Sources include: (1) 10 CFR Part 50, Appendix A, Introduction and 10 CFR

50, Appendix A, Criterion 35; (2) 10 CFR 50, Appendix K, Item I.D.1; and (3) FSAR

commitments. During an engineering review following an event, it was found that a coil

shorting in one of several supervisory relays, in conjunction with an accident, could lead

to a premature recirculation actuation signal. This could result in loss of water to the

ECCS pumps due to realignment of suction to the containment sump. It was also found

that such coil shorting could cause closure of pump recirculation isolation valves,

potentially dead heading and possibly damaging the HPSI pumps.

The licensee determined that each of these concerns represented a condition outside of

the design bases. When each determination was made, an ENS notification was made

and immediate actions were taken to fix the single failure problem. Subsequently, an

LER was submitted.

The event is reportable because the ECCS failed to meet its design bases.

(2) Design/Hardware Problem (Turbine Missile Protection)

The original design criteria, as stated in the UFSAR, required that ESFs be protected

from turbine generated missiles by means of shielding or separation. As a result of a

service water upgrade project it was found that portions of the low pressure service

water system (LPSW) did not meet the plant's separation criteria for high trajectory

turbine missiles. The licensee provided an ENS notification under 10 CFR

50.72(b)(1)(ii)(B) and submitted an LER under 10 CFR 50.73(a)(2)(ii)(B), outside design

basis.

NUREG-1022, Rev. I 38
The corrective action included submitting a UFSAR amendment, which was approved by

the NRC staff, to allow using current NRC and industry guidance. When applying this

guidance, the LPSW piping in question provides an acceptably low probability target.

This event is reportable because the turbine missile protection did not meet the design

basis as stated in the FSAR.

(3) ECCS Analysis

The large break LOCA analysis, as documented in the FSAR, assumed that high and

low head safety injection systems can deliver full flow in 5 and 10 seconds, respectively.

A new analysis was performed, which accounted conservatively for (1) SI signal

processing, (2) sequencer delay time uncertainty, and (3) increased time for pump

acceleration to full speed due to degraded voltage. This indicated that it could take as

much as 8 and 24 seconds, respectively, for the high and low head safety injection

systems to deliver full flow.

As a result of the new analysis, calculated peak clad temperature was increased by

about 44F. However, peak clad temperature remained below 2200F and other ECCS

acceptance criteria continued to be met as well. Although licensee reported the event

as outside design bases, staff does not consider the event reportable under that

criterion because the provisions of §50.46(a) apply. Under those provisions, the events

reportable pursuant to §50.72 and §50.73 are those where the ECCS acceptance

criteria are exceeded.

The event is not reportable because the provisions of §50.46(a) apply and the ECCS

acceptance criteria were not exceeded.

(4) Fire Protection (Separation of Safe Shutdown Trains)

The design for a fire area had been approved on the basis of several specific features

including: automatic sprinklers; remote annunciation; and circuit separation via 2-hour

rated fire barriers for redundant safe shutdown circuits. During a design basis review it

was found that redundant diesel generator field circuits, located in a common fire area,

were not protected or separated by 2-hour rated fire barriers.

An ENS notification was made under 10 CFR 50.72(b)(1)(ii)(B), outside design basis,

and subsequently an LER was submitted under 10 CFR 50.73(a)(2)(ii)(B), outside

design basis.

The condition was reportable under this criterion because the required design basis

protective features for safe shutdown trains, as described in 10 CFR 50, Appendix R

and the FSAR, were lacking.

39 NUREG-1022, Rev. 1
(5) Hardware Problem (Suitable Redundancy and -Seismic Qualification)

During an NRC evaluation, it was found that an exciter panel for one diesel generator

had lacked appropriate seismic restraints since the plant was constructed. The licensee

did not initially believe the condition was reportable under 10 CFR 50.72(b)(1)(ii)(B) and

10 CFR 50.73(a)(2)(ii)(B), outside the design basis. However, the staff determined that

the condition was reportable under this criterion because the onsite power system

lacked suitable redundancy (seismically qualified) as described in GDC 2, GDC 17 and

the SAR.1161

{16) The single failure criterion is discussed in 10 CFR 50 Appendix A, Criterion 17 -

Electric Power Systems and the seismic design bases are discussed in 10 CFR 50, Appendix

A, Criterion 2 - Design Bases for Protection Against Natural Phenomena, as well as in the

FSAR.

NUREG-1022, Rev. I 40
3.2.5 External Threat to Plant Safety

§50.72(b)(1)(iii) §50.73(a)(2)(iii)

Licensee shall report: "Any natural Licensee shall report: "Any natural

phenomenon or other external condition that phenomenon or other external condition that

poses an actual threat to the safety of the posed an actual threat to the safety of the

nuclear power plant or significantly hamper1 nuclear power plant or significantly

site personnel in the performance of duties hampered site personnel in the performance

necessary for the safe operation of the of duties necessary for the safe operation of

plant." the nuclear power plant."

If not reported as an emergency under §50.72(a), licensees are required to report any natural

phenomenon or other external condition that poses an actual threat to the safety of the nuclear

power plant or significantly hampers site personnel in the performance of duties necessary for

the safe operation of the plant to the NRC via the ENS as soon as practical and in all cases

within 1 hour. Subsequent evaluation may indicate that the phenomenon did not pose an actual

threat or significantly hamper site personnel. If so, an LER is not required and the ENS

notification may be retracted. Otherwise, licensees are required to submit an LER within 30

days.

Discussion

These criteria apply only to acts of nature (e.g., tornadoes, earthquakes, fires, lightning,

hurricanes, floods) and external hazards (i.e., industrial or transportation accidents).

References to acts of sabotage are covered by 10 CFR 73.71. Actual threats or significant

hampering from internal hazards are covered by separate criteria in §50.72(b)(1)(vi) and

§50.73(a)(2)(x), as discussed in Section 3.2.8 of this report.

For ENS reporting, the phrase "actual threat to safety of the nuclear power plant" is one

reporting trigger. This covers those events involving an actual threat to the plant from an

external condition or natural phenomenon where the threat or damage challenges the ability of

the plant to continue to operate in a safe manner (including the orderly shutdown and

maintenance of shutdown conditions).

The licensee should decide if a phenomenon or condition actually threatens the plant. For

example, a minor brush fire in a remote area of the site that is quickly controlled by fire fighting

personnel and, as a result, did not present a threat to the plant should not be reported.

However, a major forest fire, large-scale flood, or major earthquake that presents a clear threat

to the plant should be reported. As another example, an industrial or transportation accident

which occurs near the site, creating a plant safety concern, should be reported.

The licensee must use engineering judgment to determine if there was an actual.threat. For

example, with regard to tornadoes the decision would be based on such factors as the size of

41 NUREG-1022, Rev. I
the tornado, and its location and path. There are no prescribed limits. In general, situations

involving only monitoring by the plant's staff are not reportable, but if preventive actions are

taken or if there are serious concerns, then the situation should be carefully reviewed for

reportability.

Responsive actions, by themselves, do not necessarily indicate actual threats. Those which

are purely precautionary, such as placement of sandbags, even though flood levels are not

expected to be high enough to require sandbags, do not trigger reporting.

Some natural phenomena such as floods may be accurately predicted. If there isa credible

prediction of a flood that would challenge the ability of the plant to continue to operate safelyi

the threat is reportable as an actual threat via ENS as soon as practical and in all cases within 1.

hour.

In most cases, events such as earthquakes, approaching hurricanes or tornado warnings result

in ENS notification because there is a declaration of an emergency class, which is reportable

under §50.72(a)(1)(i) as discussed in Section 3.1.1 of this report, rather than because the event

is considered an actual threat. Usually, with the passage of time, it is apparent that an actual

threat did not occur and, thus, no LER is submitted (see Example 1). In some cases, with the

passage of time, it is judged that an actual threat did occur and, thus, an LER is submitted (see

Example 2).

Section 3.2.8 of this report discusses the meaning of the phrase "significantly hampers site

personnel in the performance of duties necessary for the safe operation of the plant," in the

context of internal threats. A natural phenomenon or external condition, may also significantly

hamper personnel. If so, it is reportable under this criterion.

If a snowstorm, hurricane or similar event significantly hampers personnel in the conduct of

activities necessary for the safe operation of the plant, the event is reportable via the ENS as

soon as practical and in all cases within 1 hour. In the case of snow, the licensee must use

judgment based on the amount of snow, the extent to which personnel were hampered, the

extent to which additional assistance could have been available in an emergency, the length of

time the condition existed, etc. For example, if snow prevented shift relief for several hours, the

situation would be reportable if the delay were such that site personnel were significantly

hampered in the performance of duties necessary for safe operation. For example, shift

personnel might exceed normal shift overtime limits, become. excessively fatigued, or find i

necessary to operate with fewer than the required number of watchstanders in order to allow

some to rest.

Examples

(1) Earthquake

Seismic alarms were received in the Unit 1 control room of a Southern California plant.

Seismic monitors were not tripped in Units 2 or 3. The earthquake was readily felt on

site. Seismic instrumentation measured less than 0.02g lateral acceleration.

NUREG-1022, Rev. I 42
The licensee classified this as an Unusual Event in accordance with the emergency plan

and notified the NRC via ENS per §50.72(a)(1)(i) within 30 minutes of the earthquake.

The licensee terminated the event after walkdowns of the plant were satisfactorily

completed and made an ENS update call. No LER was submitted because the event

was not considered to be an actual threat.

(2) Hurricane

A licensee in southern Florida declared an Unusual Event after a hurricane warning was

issued by the National Hurricane Center. The hurricane was predicted to reach the site

in approximately 24 hours. As part of the licensee's severe weather preparations both

operating units were taken to hot shutdown before the hurricane's predicted arrival.

Offsite power to both units was lost. As the hurricane approached, wind velocity on site

was measured in excess of 140 mph. All personnel were withdrawn to protected safetyrelated structures. Extensive damage occurred on site. The Unusual Event was

upgraded to an Alert when the pressurized fire header was lost because of stormrelated damage to the fire protection system water supply piping and electric pump. All

safety-related equipment functioned as designed before, during, and after the storm with

the exception of two minor emergency diesel generator anomalies. The licensee

downgraded the Alert to an Unusual Event once offsite power was restored and a

damage assessment completed.

An ENS notification was required because the licensee declared an emergency class.

The licensee submitted an LER within 30 days of the hurricane, based on the

occurrence of a natural phenomenon that posed an actual threat and several other

reporting criteria as well.

(3) Fire

With the unit at 100-percent power, the control room was notified that a forest fire was

burning west of the plant close to the 230-kV distribution lines. Approximately 15

minutes later, voltage fluctuations were observed and then a full reactor scram occurred.

The licensee determined that the offsite distribution breakers had tripped on fault,

apparently from heavy smoke and heat in the vicinity of the offsite 230-kV line

insulators. The other source of offsite power, i.e., the 34.5-kV lines supplying the

startup transformers, was also lost. Both station emergency diesel generators received

a fast start signal and load sequenced as designed. Five minutes later, offsite power

was available through the startup transformer to the non-safety-related 4160-v buses,

but the licensee decided to maintain the vital buses on their emergency power source

until the reliability of offsite power could be assured. The fire continued to bum and,

although no plant structures or equipment were directly affected, the fire did approach

within 70 feet of the fire pump house.

The licensee entered the emergency plan, declaring an Unusual Event based on high

drywell temperature and an Alert based on the potential of the forest fire to further affect

the plant. The licensee submitted an LER within 30 days of the fire, based on the

43 NUREG-1022, Rev. 1
occurrence of natural phenomenon that posed an actual threat and several other

reporting criteria as well.

NUREG-1022, Rev. 1 44
3.2.6 ECCS Discharge Into the Reactor Coolant System

§50.72(b)(1)(iv) 10 CFR 50.73

Licensees shall report: "Any event that [ECCS discharge is a subset of

results or should have resulted in §50.73(a)(2)(iv), actuation of an engineered

Emergency Core Cooling System (ECCS) safety feature (ESF), as discussed in

discharge into the reactor coolant system as Section 3.3.2. Therefore, an LER is

a result of a valid signal." required.]

If not reported as an emergency under §50.72(a), licensees are required to notify the NRC via

the ENS when a discharge of the ECCS into the RCS occurred or should have occurred as a

result of a valid signal as soon as practical and in all cases within 1 hour.

Discussion

Experience with ENS notifications has shown that events involving ECCS discharge to the

vessel are generally more serious than ESF actuations without discharge to the vessel. On the

basis of this experience, the Commission has made this reporting criterion a 1-hour report.

Those events that result in either automatic or manual actuation of the ECCS or would have

resulted in activation of the ECCS if some component had not failed or an operator action had

not been taken are reportable. For example, if a valid ECCS signal was generated by plant

conditions and the operator put all ECCS pumps in pull-to-lock position, although no ECCS

discharge occurred, the event is reportable.

A "valid signal" refers to the actual plant conditions or parameters satisfying the requirements

for ECCS initiation. Valid actuations also include intentional manual actuations, unless the

actuation is part of a preplanned sequence during test or operation. Excluded from this

reporting requirement would be those instances in which instrument drift, spurious signals,

human error, or other invalid signals caused actuation of the ECCS (e.g., jarring a cabinet, an

error in the use of jumpers or lifted leads, an error in the actuation of switches or controls,

equipment failure or radio frequency interference). However, such events may be reportable

under other criteria; in particular, if an ESF is actuated §50.72(b)(2)(ii) requires a report within

four hours and §50.73(a)(2)(iv) requires submittal of an LER.

The staff considers deliberate manual ECCS initiations or actuations based on the operator's

understanding of actual plant conditions or parameters as valid signals. However, inadvertent

manual ECCS initiations or actuations that occur because of human error, such as errors that

occur during surveillance tests or maintenance activities, are not considered as valid signals. If

the ECCS discharged or should have discharged into the reactor coolant system as a result of

an invalid signal, no ENS notification under this reporting criterion is required. (Such a condition

may be reportable as an ESF actuation under 10 CFR 50.72(b)(2)(ii).)

Any event reportable under §50.72(b)(1)(iv) also requires a 30-day LER under §50.73(a)(2)(iv)

because an ESF was actuated.

45 NUREG-1022, Rev. 1
Examples

(1) BWR Scram and ECCS Injection on Valid Signal

A loss of instrument air caused the feedwater pump minimum flow valves to fail open

and decrease reactor vessel level. This resulted in an automatic reactor scram/turbine

trip and high-pressure core spray and reactor core isolation cooling injection into the

reactor vessel for 4 minutes. After reactor vessel level and the condensate and

feedwater systems were restored, these pumps were secured.

An ENS notification is required under §50.72(b)(1)(iv) because an ECCS system

injected water into the RCS as a result of a valid ECCS signal. Although the RPS

actuation also is reportable within 4 hours under §50.72(b)(2)(ii), this more limiting

criterion applies. An LER is required under §50.73(a)(2)(iv) because an ESF actuation

occurred.

(2) PWR ECCS Injection following Surveillance Testing

While making preparations for a normal plant cooldown in Mode 5, the licensee

performed stroke time testing of the safety injection isolation valves. Following the test

these valves were not returned to the closed position. This resulted in approximately

2000 gallons of borated water injecting into the reactor coolant system when the plant

was depressurized below the safety injection tank pressure of 260 psia.

This event is reportable as an ECCS injection under §50.72(b)(1)(iv). ECCS initiation

was based on RCS pressure being less than safety injection tank pressure. Therefore,

ECCS initiation is considered to result from a valid signal. An LER is required under

§50.73(a)(2)(iv).

(3) PWR ECCS Injection Caused by Personnel Error

While surveillance testing containment isolation valves, a test push-button was

inadvertently released, which initiated a "B" train containment isolation and ECCS.

High-pressure ECCS pumps injected 300 gallons of borated water from the refueling

water storage tank into the reactor before the "B" pumps were secured while the reactor

remained at 94-percent power.

This event is not reportable under §50.72(b)(1)(iv), even though it was an ECCS

injection into the RCS, because it resulted from an invalid signal; however, it is

reportable as an ESF actuation under §50.72(b)(2)(ii) and an LER is required under

§50.73(a)(2)(iv).

NUREG-1022, Rev. 1 46
3.2.7 Loss of Emergency Preparedness Capabilities

§50.72(b)(1)(v) 110 CFR 50.73

Licensees shall report: "Any event that [No corresponding Part 50.73 requirement.]

results in a major loss of emergency

assessment capability, offsite response

capability, or communications capability

(e.g., significant portion of control room

indication, Emergency Notification System,

or offsite notification system)."

If not reported as an emergency under 50.72(a), licensees are required to notify the NRC of a

major loss of their emergency assessment, offsite response, or communications capability as

soon as practical and in all cases within 1 hour.

Discussion

This reporting requirement pertains to events that would impair a licensee's ability to deal with

an accident or emergency. Notifying the NRC of these events may permit the NRC to take

some compensating measures and to more completely assess the consequences of such a

loss should it occur during an accident or emergency.

Examples of events that this criterion is intended to cover are those in which any of the

following is not available:

* Safety parameter display system (SPDS)

• Emergency response facilities (ERFs)

Emergency communications facilities and equipment including the emergency

notification system (ENS)

• Public prompt notification system including sirens

* Plant monitors necessary for accident assessment

These and other situations should be evaluated for reportability as discussed below.

Loss of Emergency Assessment Capability

A major loss of emergency assessment capability would include those events that significantly

impair the licensee's safety assessment capability. Some engineering judgment is needed to

determine the significance of the loss of particular equipment, e.g., loss of only the SPDS for a

47 NUREG-1022, Rev. 1
short period of time need not be reported, but loss of SPDS and other assessment equipment

at the same time may be reportable.

The staff considers the loss of a significant portion of control room indication including

annunciators or monitors, or the loss of all plant vent stack radiation monitors, as examples of a

major loss of emergency assessment capability which should be evaluated for reportability.

Loss of Offsite Response Capability

A major loss of offsite response capability includes those events that would significantly impair

the fulfillment of the licensee's approved emergency plan for other than a short time. Loss of

offsite response capability may typically include the loss of plant access, emergency offsite

response facilitiesc'1, or public prompt notification system, including sirens and other alerting

systems.

If a forge-tm significaa natural hazard.(e.g., earthquakehuricanetomado, flood, etc,) or

other event causes reads to be elmesed eva~cuation routes to be impassible or other parts of the

response infrastructure to benimpaired to the extent that the State and local governments are

rendered incapable of fulfilling their responsibilities in the emergency plan for the plant, then the

NRC must be notified. This doenot apply in th..case ofroutine.traffic impedimentssuch as

fog, snow and ice which do not- render the state and local governments incapable of fulfilling

their responsibilities. It is intended to apply to more significant cases such as the conditionsi

around the Turkey Point plant after Hurricane Andrew struck in 19992 or the conditions anund

the Cooper station during the Midwest floods of 1993.

If the alert systems, e.g., sirens, are owned and/or maintained by others, the licensee should

take reasonable measures to remain informed and must notify the NRC if a large number of

sirens fail. Although the loss of a single siren for a short time is not a major loss of offsite

response capability, the loss of a large number of sirens, other alerting systems (e.g., tone alert

radios), or more importantly, the lost capability to alert a large segment of the population for 1

hour would warrant an immediate notification.

Loss of Communications Capability

A major loss of communications capability may include the loss of ENS and/or other offsite

communication systems. The other offsite communication systems may include a dedicated

telephone communication link to a State or a local government agency and emergency offsite

response facilities, in-plant paging and radio systems required for safe plant operation, or

commercial telephone lines.

Should either or both of the emergency communications subsystems (ENS and HPN) fail, the

NRC Operations Center should be so informed over normal commercial telephone lines. When

notifying the NRC Operations Center, licensees should use the backup commercial telephone

[')Performing preventive Maintenance on an offsite emergency response facility Is not

reportable if the facility can be returned to service promptly in the event of an accident.

NUREG-1022, Rev. 1 48
numbers provided. This satisfies the guidance provided in previous Information Notices 85-44

"Emergency Communication System Monthly Test," dated May 30, 1985 and 86-97

"Emergency Communications System," dated November 28, 1986, to test the backup means of

communication when the primary system is unavailable as well as the reporting requirements of

§50.72(b)(1)(v). If the Operations Center notifies the licensee that an ENS line is inoperable,

there is no need for a subsequent licensee notification. Loss of either ENS or HPN does not

generate an event report. The Operations Center contacts the appropriate repair organization.

In a similar man~neirif the flRC supplied telephone line or modem used for the emergency

response data system is inoperable, the NRC operations center should be informied s. that

repa~irs ca~n be orde~red. However, this does not generate an event report.

Examples

(1) Plant Access Roads Closed by Storm

The local sheriff notified the licensee that all roads to and from the plant were closed

because of a snow storm. The licensee had two full-shift crews on site to support plant

operations and no emergency declaration was made. The licensee notified State and

local authorities of the situation and made an ENS notification. The licensee deactivated

its station isolation procedures after the storm passed and the roads were passable.

An ENS notification was made because the licensee determined that the road closing

constituted a major loss of emergency offsite response capability. No LER is required.

(2) Loss of Public Prompt Notification System

ENS notifications of the loss of the emergency sirens or tone alert radios vary according

to the licensee's locale and interpretations of "major loss" and have included:

* 12 of 40 county alert sirens disabled because of loss of power as a result of

severe weather.

* 28 of 54 alert sirens were reported out of service as a result of a local ice storm.

* All offsite emergency sirens were:

- found inoperable during a monthly test:

- taken out of service for repair.

- inoperable because control panel power was lost.

- inoperable because the county radio transmitter failed.

An ENS notification is required because of the major loss of offsite response capability,

i.e., the public prompt notification system. However, licensees may use engineering

judgment in determining reportability (i.e., a "major loss") based upon such factors as

the percent of the population not covered by emergency sirens and the existence of

49 NUREG-1022, Rev. 1
procedures or practices to compensate for the lost emergency sirens. An LER is not

required because there is no corresponding 10 CFR 50.73 requirements.

(3) Loss of ENS and Commercial Telephone System

The licensee determined that ENS and commercial telecommunications capability was

lost to the control room when a fiber optic cable was severed during maintenance. A

communications link was established and maintained between the site and the load

dispatcher via microwave transmission. Both the ENS and commercial communications

capability were restored approximately 90 minutes later.

An ENS notification is required because of the major loss of communications capability.

Although the microwave link to the site was established and maintained during the

telephone outage, this in itself does not fully compensate for the loss of communication

that would be required in the event of an emergency at the plant. No LER is required

because there is no corresponding 10 CFR 50.73 requirements.

(4) Loss of Direct Communication Line to Police

The licensee contacted the State Police via commercial telephone lines and reported to

the NRC Operations Center that the direct telephone line to the State Police was

inoperable for over 1 hour. The licensee notified the NRC Operations Center in a

followup ENS call that the line was restored to operability.

An ENS notification would be required if the loss of the direct telephone line(s) to various

police, local, or State emergency or regulatory agencies is not compensated for by other

readily available offsite communications systems. In this example, no ENS notification

is required since commercial telephone lines to the State Police were available. No LER

is required because there is no corresponding 10 CFR 50.73 requirements.

NUREG-1022, Rev. 1 50
3.2.8 Internal Threat to Plant Safety

§50.72(b)(1)(vi) §50.73(a)(2)(x)

Licensees shall report: "Any event that Licensees shall report: "Any event that

poses an actual threat to the safety of the posed an actual threat to the safety of the

nuclear power plant or significantly hamperl nuclear power plant or significantly

site personnel in the performance of duties hampered site personnel in the performance

necessary for the safe operation of the of duties necessary for the safe operation of

nuclear power plant including fires, toxic gas the nuclear power plant including fires, toxic

releases, or radioactive releases." gas releases, or radioactive releases."

If not reported as an emergency under §50.72(a), licensees are required to report such an

event or condition to the NRC via the ENS as soon as practical and in all cases within 1 hour.

Licensees are required to submit an LER within 30 days.

Discussion

These criteria pertain to internal threats. The criteria for external threats, §50.72(b)(1)(iii) and

§50.73(a)(2)(iii), are described in Section 3.2.5.

This provision requires reporting events, particularly those caused by acts of personnel, which

endanger the safety of the plant or interfere with personnel in the performance of duties

necessary for safe plant operations.

The licensee must exercise some judgment in reporting under this rule. For example, a small

fire on site that did not endanger any plant equipment and did not and could not reasonably be

expected to endanger the plant is not reportable.

As indicated in the Statement of Considerations the phrase "significantly hampers site

personnel" applies narrowly, i.e. only to those events which significantly hamper the ability of

site personnel to perform safety-related activities affecting plant safety.01 8)

In addition, the staff considers the following standards appropriate in this regard:

The significant hampering criterion is pertinent to "the performance of duties necessary

for safe operation of the nuclear power plant." One way to evaluate this i• to asikif one

tould sele the-room ki question (or disable' the function in question) for a substantial

period of time an.d.s~till ope~rate -th~e plant sa~fely, Fo~rexample. if a switchgear room it

Unavailable for a time, but it is normally not necessary to enter the room far safe

operation, and..no .need..toenter the room, arises while it is unavailable, the event is not

portable uncler this criterion.

{18) 48 FR 33856, July 26, 1983.

51 NUREG-1022, Rev. 1
* Significant hampering includes hindering or interfering (such as with protective clothing

or radiation work permits) provided that the interference or delay is sufficient to

significantly threaten the safe operation of the plant.

* Actions such as room evacuations that are precautionary would not constitute significant

hampering if the necessary actions can still be performed in a timely manner.

Plant mode may be considered in determining if there is an actual internal threat to a plant.

However, licensees should not incorrectly assume that everything that happens while a plant is

shut down is unimportant and not reportable.

In-plant releases must be reported if they require evacuation of rooms or buildings eentainrig

.y.t. "- impert-- t to safety and, as a result, the ability of the operators to perform necessary

duties is significantly hampered.

Events such as minor spills, small gaseous waste releases, or the disturbance of contaminated

particulate matter (e.g., dust) that require temporary evacuation of an individual room until the

airborne concentrations decrease or until respiratory protection devices are used, are not

reportable unless the ability of site personnel to perform necessary safety functions is

significantly hampered.

No LER is required for precautionary evacuations of rooms and buildings that subsequent

evaluation determines were not required. Even if an evacuation affects a major part of the

facility, the test for reportability is whether an actual threat to plant safety occurred or whether

site personnel were significantly hampered in carrying out their safety responsibilities.

In most cases, fires result in ENS notification because there is a declaration of an emergency

class, which is reportable under §50.72(a)(1)(ii) as discussed in Section 3.1.1 of this report.'19)

If there is an actual threat or significant hampering, an LER is also required. With regard to

control room fires, the staff generally considers a control room fire to constitute an actual threat

and significant hampering.120'

(19) As indicated in NUREG-0654, Rev. 1, Information Notice 88-64 and Regulatory

Guide 1.101, Rev. 3 (which endorses NUMARC/NESP-007, Rev. 2), a fire that lasts longer than

10 or 15 minutes or which affects plant equipment important for safe operation would result in

declaration of an emergency class.

It is theoretically possible to have a control room fire which is discovered and

extinguished quickly and, even in this location, does not significantly hamper the operators and

does not threaten plant safety. Examples could include small paper fires in ash trays or trash

cans, or cigarette bums of furniture or upholstery.

NUREG-1022, Rev. 1 52
Examples

(1) Fires

* Question:

If we have a fire in the refueling bridge and we are not moving fuel, would the fire be

reportable?

Answer:

No. If the plant is not moving fuel and the fire does not otherwise threaten other safety

equipment and does not hamper site personnel, the fire is not reportable. If the plant is

moving fuel, the fire is reportable.

* Question:

If we have a fire in the reactor building that forces contractor personnel who are doing a

safety related modification to leave, but the fire did not hamper operations personnel or

equipment, would that fire be reportable?

Answer:

No. The fire would not be reportable if the fire was not severe enough that it posed an

actual threat to the plant and the delay in completing the modification did not

significantly threaten the safe operation of the plant.

53 NUREG-1022, Rev. I
3.3 Four-hour ENS Notifications and LERs

This section addresses §50.72(b)(2), "Non-Emergency Events--Four-Hour Reports," and 10

CFR 50.73 written reports associated with these 50.72 notifications. If not reported as a

declaration of emergency class under §50.72(a) or as a non-emergency 1-hour report under

§50.72(b)(1), licensees are to notify the NRC as soon as practical and in all cases within 4

hours of the occurrence of any of the events required by §50.72(b)(2) and to submit an LER

within 30 days for any event or condition required by 10 CFR 50.73.

In addition to events reportable under both 10 CFR 50.72 and 50.73, several requirements for

50.72 notifications only or LERs only are included in this section because of the sequential

numbering scheme used. For example, common-mode failures of channels, trains, or systems,

as discussed in Section 3.3.4, require LERs, but no ENS notifications are explicitly required

unless reportable under other criteria. Transport of a contaminated person to an offsite medical

facility, as discussed in Section 3.3.6, requires ENS notification but no LER.

NUREG-1022, Rev. I 54
3.3.1 Shutdown Plant Found in Degraded or Unanalyzed Condition

§50.72(b)(2)(i) 10 CFR 50.73

Licensees shall report: "Any event found [Events found while the reactor is shutdown

while the reactor is shut down. that, badit that involve degradation of the principal

been found while the reactor was in safety barriers or unanalyzed conditions that

operation, would have resulted in the significantly compromise plant safety are

nuclear power plant, including its principal addressed by §50.73(a)(2)(ii). Therefore, an

safety barriers, being seriously degraded or LER is required. See Section 3.2.4.]

being in an unanalyzed condition that

significantly compromises plant safety."

If not reported under §50.72(a) or (b)(1), licensees are required to report any such condition to

the NRC via the ENS as soon as practical, and in all cases within 4 hours of discovery of the

condition. Licensees are required to submit an LER within 30 days.

Discussion

Guidelines for identifying events that would result in the nuclear power plant being seriously

degraded or being in an unanalyzed condition that significantly compromises plant safety are

discussed in Section 3.2.4 of this report.

Examples

(1) Significant Degradation of Reactor Fuel Rod Cladding Identified During Testing of Fuel

Assemblies

Radio-chemistry data for a particular PWR indicated that a number of fuel rods had

failed during the first few months of operation. Projections ranged from 6 to 12 failed

rods. The end of cycle reactor coolant system iodine-131 activity averaged 0.025 micro

curies per milliliter. following the end of cycle shutdown, iodine-1 31 spiked to 11.45

micro curies per milliliter. The cause was due to a significant number of failed fuel rods.

Inspections revealed that 136 of the total 157 fuel assemblies contained failed fuel

(approximately 300 fuel rods had through-wall penetrations), far exceeding the

anticipated number of failures. The defects were generally pinhole sized. The fuel

cladding failures were caused by long-term fretting from debris that became lodged

between the lower fuel assembly nozzle and the first spacer grid, resulting in penetration

of the stainless-steel fuel cladding. The source of the debris was apparently a

machining byproduct from the thermal shield support system repairs during the previous

refueling outage.

An ENS notification is required because a principal safety barrier (the fuel cladding) was

found seriously degraded. An LER is required.

55 NUREG-1022, Rev. 1
(2) Corrosion of a Control Rod Drive Mechanism Flange Resulted in a Reactor Coolant

System Pressure Boundary Degradation

While the plant was in hot shutdown, a total of six control rod drive mechanism (CRDM)

reactor vessel nozzle flanges were identified as leaking. Subsequently one of the

flanges was found eroded and pitted. While removing the nut ring from beneath the

flange, it was discovered that approximately 50 percent of one of the nut ring halves had

corroded away and that two of the four bolt holes in the corroded nut ring half were

degraded to the point where there was no bolt/thread engagement.

An inspection of the flanges and spiral wound gaskets, which were removed from

between the flanges, revealed that the cause of the leaks was the gradual deterioration

of the gaskets from age. A replacement CRDM was installed and the gaskets on all six

CRDMs were replaced with new design graphite-type gaskets.

An ENS notification is required because the condition caused a significant degradation

of the RCS pressure boundary. An LER is required.

(3) Significant Degradation of Reactor Fuel Rod Cladding Identified During Fuel Sipping

Operations

With the plant in cold shutdown, fuel sipping operations identified a significant portion of

cycle 2 fuel, type "LYP," had failed, i.e., four confirmed and twelve potential fuel leakers.

The potential fuel leakers had only been sipped once prior to making the ENS

notification. The licensee contacted the fuel vendor for assistance on-site in evaluating

this problem.

As in example (1), an ENS notification was made because a principal safety barrier (the

fuel cladding) was found seriously degraded. However, additional sipping operations

and a subsequent evaluation by the licensee's reactor engineering department with

vendor assistance concluded that no additional fuel failures had occurred, i.e., the

abnormal readings associated with the potential fuel leakers was attributed to fission

products trapped in the crud layer. Based on the results of the evaluation the licensee

concluded that the fuel cladding was not seriously degraded and that the event was not

reportable. Consequently, after discussion with the Regional Office, the licensee

retracted this event.

NUREG-1022, Rev. 1 56
3.3.2 Actuation of an Engineered Safety Feature or the RPS

§50.72(b)(2)(ii)

Licensees shall report "any event or

condition that results in a manual or

automatic actuation of any Engineered

Safety Feature (ESF), including the Reactor

Protection System (RPS) except when:

(A) The actuation results from and is part of

the preplanned sequence during testing or

reactor operation;

(B) The actuation ji invalid and:

(1) Occura while the system is

properly removed from service;

(2) Occurs after the safety function has

been already completed; or

(3) Involves only the following specific

ESFs or their equivalent systems;

(i) Reactor water clean-up system;

(ii) Control room emergency ventilation

system;

(iii) Reactor building ventilation system;

(iv) Fuel building ventilation system; or

(v) Auxiliary building ventilation system."

§50.73(a)(2)(iv)

Licensees shall report "any event or

condition that resulted in a manual or

automatic actuation of any Engineered

Safety Feature (ESF), including the Reactor

Protection System (RPS), except when:

(A) The actuation resulted from and was

part of a pre-planned sequence during

testing or reactor operation;

(B) The actuation was invalid and:

(1) Occurred while the system was

properly removed from service;

(2) Occurred after the safety function had

been already completed; or

(3) Involved only the following specific

ESFs or their equivalent systems;

(i) Reactor water clean-up system;

(ii) Control room emergency ventilation

system;

(iii) Reactor building ventilation system;

(iv) Fuel building ventilation system; or

(v) Auxiliary building ventilation system."

If not reported under §50.72(a) or (b)(1), licensees are required to report any engineered safety

feature actuation, including the reactor protection system, to the NRC via the ENS as soon as

practical and in all cases within 4 hours of the event. Licensees are required to submit an LER

within 30 days.

Discussion

The Statements of Considerations indicate that this paragraph requires events to be reported

whenever an ESF actuates either manually or automatically, regardless of plant status. It is

based on the premise that the ESFs are provided to mitigate the consequences of a significant

event and, therefore: (1) they should work properly when called upon, and (2) they should not

be challenged frequently or unnecessarily. The Commission is interested both in events where

an ESF was needed to mitigate the consequences (whether or not the equipment performed

properly) and events where an ESF actuated unnecessarily. In discussing the reporting of

actuations which are part of preplanned procedures, the Statements of Considerations also

state that actuations that need not be reported are those initiated for reasons other than to

57 NUREG-1022, Rev. I
mitigate the consequences of an event (e.g., at the discretion of the licensee as part of a

preplanned procedure).(21)

This indicates an intent to require reporting actuations of features that mitigate the

consequences of significant events. Usually, the staff would not consider this to include single

component actuations because single components of complex systems, by themselves, usually

do not mitigate the consequences of significant events, However, in some cases a component

would be sufficient to mitigate the event (i.e., perform the ESF function) and i actuation would,

therefore. be reportable. -This position is consistent with the statement that the reporting

requirement is based on the premise that ESFs are provided to mitigate the consequences of a

significant event:LSingle trains do mitigate the consequences, and, thus, train level actuations are reportable.

In this regard, the staff considers actuation of a diesel-generator tobe actuation of a train-not

actuation of a single component - because a diesel generator mitigates the event (performs the

ESF function forplants at which diesel generators are classified as ESF systems). (See

Example 3 below.)

The staff also considers intentional manual actions, in which one or more ESF components are

actuated in response to actual plant conditions resulting from equipment failure or human error,

to be reportable because such actions would usually mitigate the consequences of a significant

event. This position is consistent with thelstatement that the Commission is interested in

events where an ESF was needed to mitigate the consequences of the event. For example,

starting a safety injection pump in response to a rapidly deceasing pressurizer level or starting

HPCI in response to a loss of feedcwater would be reportable. However, shifting alignment of

makeup pumps or closing a containment isolation valve for normal operational purposes would

not be: reportable.

The Statement of Considerations also indicates that "actuation" of multichannel ESF actuation

systems is defined as actuation of enough channels to complete the minimum actuation logic.

Therefore, single channel actuations, whether caused by failures or otherwise, are not

reportable if they do not complete the minimum actuation logic.(22) Note, howevyerthat if only a

single logic channel, actuates when, in fact, the ESF system should have actuated in response

to plant parameters, this would be reportable as an ESF failure. The event would be reportable

under these criteria (ESF actuation) as well as under 10 CFR 50.72(b)(2)Qii) and 10 CFR

50.73(a)(2)(v) (event or condition alone). This position is. consistent with the statement that the

Commission is interested in events where an ESF was needed to mitigate the consequences,

whether or not the equipment performed properly.'23

(21) 48 FR 33854, July 28, 1983, 48 FR 39043 and 48 FR 39044, August 29, 1983.

(22) 48 FR 33854, July 28, 1983, 48 FR 39043 and 48 FR 39044, August 29, 1983.

{23) Also see 48 FR 39043, August 29, 1983, which states that this paragraph is

intended to capture events during which an ESF actuates or fails to actuate.

NUREG-1022, Rev. I 58
With regard to preplanned actuations, the Statements of Consideration indicate that operation

of an ESF as part of a planned test or operational evolution need not be reported. Preplanned

actuations are those which are expected to actually occur due to preplanned activities covered

by procedures. Such actuations are those for which a procedural step or other appropriate

documentation indicates the specific ESF actuation that is actually expected to occur. Control

room personnel are aware of the specific signal generation before its occurrence or indication in

the control room. However, if during the test or evolution, the ESF actuates in a way that is not

part of the planned evolution, that actuation should be reported. For example, if the normal

reactor shutdown procedure requires that the control rods be inserted by a manual reactor

scram, the reactor scram need not be reported. However, if unanticipated conditions develop

during the shutdown that cause an automatic reactor scram, such a reactor scram should be

reported. The fact that the safety analysis assumes that an ESF will actuate automatically

during an event doe3 not eliminate the need to report that actuation. Actuations that need not

be reported are those initiated for reasons other than to mitigate the consequences of an event

(e.g., at the discretion of the licensee as part of a planned evolution).(24 )

Note that if an operator were to manually scram the reactor in anticipation of receiving ani

automatic reactor scram, this would be reportable just as the automatic scram would be

reportable.

On September 10, 1992, the Commission published final amendments to 10 CFR 50.72 and

50.73 that apply to reporting of ESF actuations. Three categories of •iid ESF actuations are

not reportable. These three categories are invalid ESF actuations of (1) systems which had

been properly removed from service, or (2) systems for which the safety function which the ESF

is intended to accomplish had already been accomplished, and (3) several specific systems

listed below.

Valid ESF actuations are those actuations that result from "valid signals" or from intentional

manual initiation, unless it is part of a preplanned test. Valid signals are those signals that are

initiated in response to actual plant conditions or parameters satisfying the requirements for

ESF initiation. Note this definition of "valid" requires that the initiation signal must be an ESF

signal. This distinction eliminates actuations which are the result of non-ESF signals from the

class of valid actuations. Invalid actuations are, by definition, those that do not meet the criteria

for being valid. Thus, invalid actuations include actuations that are not the result of valid signals

and are not intentional manual actuations.

Invalid ESF actuations that occur when the system is already properly removed from service

are not reportable if all requirements of plant procedures for removing equipment from service

have been met. This includes required clearance documentation, equipment and control board

tagging, and properly positioned valves and power supply breakers. In addition, invalid ESF

actuations that occur after the safety function has already been completed are not reportable

An example would be RPS actuation after the control rods have already been inserted into the

core.

(24) 48 FR 33854, July 28, 1983, 48 FR 39043 and 48 FR 39044, August 29, 1983.

59 NUREG-1022, Rev. 1
Finally, invalid actuations for several specific systems or their equivalent are not reportable.

These systems are the reactor water clean up system in boiling water reactors (BWRs), the

control room emergency ventilation system, the reactor building ventilation system (RBVS), the

fuel building ventilation system, and the auxiliary building ventilation system. Thus, reporting of

invalid actuations for these specific systems due to signals that originated from non-ESF

circuitry are not required.

Invalid actuations of other ESF systems continue to be reportable. For BWRs, the actuation of

the standby gas treatment system following an invalid actuation of the RBVS is also not

reportable.

If an invalid ESF actuation reveals a defect in the ESF system so the system failed or would fail

to perform its intended function, the event continues to be reportable under other requirements

of 10 CFR 50.72 and 50.73. When invalid ESF actuations excluded by the conditions described

above occur as part of a reportable event, they should be described as part of the reportable

event, in order to provide a complete, accurate and thorough description of the event.

The reporting criterion "is based on the premise that ESFs are provided to mitigate the

consequences of a significant event ... Systems typically reported under this criterion

include the systems listed in Table 2. These are systems required to: mitigate significant events

and include ECCS, RPS, containment systems and certain auxiliary and support systems

required to perform ESF functions. These are systems that are described in the FSAR and are

required to satisfy ESF functional requirements.7 The NRC staff considers these systems to be

a reasonable interpretation of what constitutes systems "provided to mitigate the consequences

of a significant event."

Examples

(1) RPS Actuation

The licensee was placing the residual heat removal (RHR) system in its

shutdown cooling mode while the plant was in hot shutdown. The BWR

vessel level decreased for unknown reasons, causing a RPS scram and

Group III primary containment isolation signals, as designed. All control

rods had been previously inserted and all Group III isolation valves had

been manually isolated. The licensee isolated RHR to stop the decrease

in reactor vessel level.

This event is reportable within 4 hours under this criterion because, although the

systems' safety functions had already been completed, the RPS scram and

primary containment isolation signals were valid and the actuations were not part

of the planned procedure. The automatic signals were valid because they were

generated from the sensor by measurement of an actual physical system

parameter that was at its set point. An LER is required.

(25) 48 FR 33854, July 26, 1983.

NUREG-1022, Rev. 1 60
* With the BWR defueled, an invalid signal actuated the RPS. There was no

component operation because the control rod drive system had been properly

removed from service. This event is not reportable because (1) the RPS signal

was invalid, and (2) the system had been properly removed from service.

• An immediate notification (§50.72) was received from a BWR licensee. In the

reported event, both recirculation pumps tripped as a result of a breaker

problem. This placed the plant in a condition in which BWRs are generally

scrammed to avoid potential power/flow oscillations. At this plant, for this

condition, a written off-normal procedure required the plant operations staff to

scram the reactor. The plant staff performed a reactor scram which was

uncomplicated. This event is reportable as a manual RPS actuation. Even

though the reactor scram was in response to an existing written procedure, this

event does not involve a preplanned sequence because the loss of recirculation

pumps and the resultant off-normal procedure entry were event driven, not

preplanned. An LER is required. In this case, the licensee initially retracted the

ENS notification believing that the event was not reportable. After staff review

and further discussion, it was agreed that the event is reportable for the reasons

discussed above.

(2) BWR Control Rod Block Monitor Actuation

A rod block that was part of the planned startup procedure occurred from the rod block

monitor, which, at this plant, is classified as a portion of the RPS or as an ESF.

ThIs vnt isrnot'reportable~becasl t oc~acurred as.'p-arto.fpr'e'piennad tartup

procedure that specrtied certa~in rod blocks were expected to occur.

(3) Emergency Diesel Generator (EDG) Starts

* The licensee provided an LER describing an event in which the EDG

automatically started when a technician inadvertently caused a short circuit that

de-energized an essential bus during a calibration. An ENS notification and LER

are required because the EDG auto-start (ESF actuation at this plant) was not

identified at the step in the calibration procedure being used.

* The licensee provided an LER describing an event in which, after an automatic

EDG start, and for unknown reasons, the emergency bus feeder breaker from

the EDG did not close when power was lost on the bus. An ENS notification and

LER are required because the actuation logic for the EDG start (ESF actuation

at this plant) was completed, even though the diesel generator did not power the

safety buses.

61 NUREG-1022, Rev. 1
(4) Preplanned Manual Scram

During a normal reactor shutdown, the reactor shutdown procedure required that reactor

power be reduced to a low power at which point the control rods were to be inserted by

a manual reactor scram. The rods were manually scrammed.

This event is not reportable because the manual scram results from and is, by

procedure, part of a preplanned sequence of reactor operation. However, if conditions

develop during the process of shutting down that require an unplanned reactor scram,

the RPS actuation (whether manually or automatically produced) is reportable via ENS

notification and LER.

(5) Actuation of Wrong Component During Testing

During surveillance testing of the main steam isolation valves (MSIVs), an

operator incorrectly closed MSIV "D" when the procedure specified closing MSIV

IIC."

This event is not reportabte because the event is an inadvertent. actuation of a single

component of an ESF system rather than a train level actuation (and the purpose of the

actuation was not to mitigate the consequences of an event);

(6) Control Room Ventilation System (CRVS) Isolation

While the CRVS was in service with no testing or maintenance in progress, a

voltage transient caused spiking of a radiation monitor resulting in isolation of the

CRVS, as designed.

This event is not reportable under this criterion because the event is due to an

invalid signal and involves one of the four excepted systems (CRVS).

(7) Reactor Water Cleanup (RWCU) Isolations

* The RWCU isolation valves closed in response to high water temperature, as

designed. This is a common operational occurrence not indicative of a

significant event; the initiation signal for this isolation is a non-ESF signal. As

discussed above, this is an invalid actuation because it originates from a nonESF signal and the event is not reportable because it is an invalid actuation of

one of the four excepted systems.

* An RWCU primary containment isolation (ESF actuation) occurred on

pressurization between the RWCU suction containment isolation valves during

the restoration of the RWCU system after a maintenance outage. An ENS

notification and LER are required because a valid ESF signal initiated the RWCU

isolation and the actuation was not part of a planned procedure.

NUREG-1022, Rev. 1 62
(8) Manual Actuation of ESF Component in Response to Actual Plant Condition

At a PWR, maintenance personnel inadvertently pulled an instrument line out of a

compression fitting connection at a pressure transmitter. The resultant reactor coolant

system (RCS) leak was estimated at between 70 and 80 gpm. Charging flow increased

due to automatic control system action. The operations staff recognized the symptoms

of an RCS leak and entered the appropriate off-normal procedure. The procedure

directed the operations staff to start a second charging pump and flow was manually

increased to raise pressurizer level. Based on the response of the pressurizer level, the

operations staff determined that a reactor scram and safety injection were not

necessary. Maintenance personnel still at the transmitter closed the instrument block

and root valves terminating the event.

The staff considers the manual start of the charging pump (which also serves as an

ECC$ pumrp, but with a different valve lineup) in response to dropp~ing pressurizer level

to be an intentional manual actuation of an ESF in response to equipment failure or

human error and reportable because it constitutes deliberate manual actuation of a

single component of an ESF. in response to plant conditions, to mitigate the

consequences of an event. As indicated in the Statements of Considerations for the

rules "Actuations that need not be reported are those that are initiated for reasons other

than to mitigate th~eco~nseque~nces o~fan ev~en. (e.g.., at the iscre~tion o~f theflice~nsee as

part of a planned procedure or evolution).?'26

(9) ESF Actuation During Maintenance Activity

At a BWR, a maintenance activity was under way involving placement of a

jumper to avoid ESF actuations. The maintenance staff recognized that there

was a high potential for a loss of contact with the jumper and consequent ESF

actuation. This potential was explicitly stated in the maintenance work request

and on a risk evaluation sheet. The operating staff was briefed on the potential

ESF actuations prior to start of work. During the event, a loss of continuity did

occur and the ESF actuations involving isolation, standby gas treatment start,

closing of some valves in the primary containment isolation system (recirculation

pump seal mini-purge valve, nitrogen supply to drywell valve, and containment

atmospheric monitoring valve) occurred.

The staff has concluded thatthe ev2ent would not4be reportable if the event were

described in appropriate documentation as definitely expected to occur. However, since

,the event was not listed as d~efinitely.e~xpe~cted to occur and was not an intended result

oDf the planned procedure, the event is reportable.

P6 48 FR 39043, August 29, 1983, and 48 FR 33854, July 26, 1983.

63 NUREG-1022, Rev. 1
Table 2. Example Systems

Emergency Core Cooling Systems (ECCS) for Pressurized Water Reactors (PWRs):

* reactor coolant system accumulators

* boron injection system

* high-, intermediate-, and low-head injection systems, including systems for

charging using centrifugal charging pumps, safety injection systems, and

residual (decay) heat removal systems

ECCS for Boiling Water Reactors (BWRs):

• high- and low-pressure core spray systems

* high-pressure coolant injection system, feedwater coolant injection system,

residual heat removal system (low pressure injection portion)

* isolation condenser system, reactor core isolation cooling system

* automatic depressurization system

Containment Systems

* containment and reactor vessel isolation systems

* containment heat removal and depressurization systems, including the

containment spray and additive system and the fan cooler system

* containment air purification and cleanup systems

* containment combustible gas control systems, including hydrogen

recombiners, igniters, and containment atmospheric dilution systems

• BWR standby gas treatment systems

Electrical Systems

* emergency ac electrical power systems, including emergency diesel

generators (EDGs) and their associated support systems and BWR dedicated

Division 3 EDGs and their associated support systems

* actuation and control systems

Heating, Ventilating and Air Conditioning (HVAC) Systems for Control Room

and Fuel Handling Areas

Anticipated Transient Without Scram (ATWS) Mitigating Systems

PWR Auxiliary Feedwater Systems

NUREG-1022, Rev. 1 64
3.3.3 Event or Condition That Alone Could Prevent Fulfillment of a Safety Function

1*

§50.72(b)(2)(iii)

Licensees shall report: "Any event or

condition that alone could have prevented

the fulfillment of the safety function of

structures or systems that are needed to:

(A) Shut down the reactor and maintain it

in a safe shutdown condition;

(B) Remove residual heat;

(C) Control the release of radioactive

material; or

(D) Mitigate the consequences of an

accident."

10 CFR 50.72

[The Statements of Consideration for 10

CFR 50.72 contain wording similar to those

of §50.73(a)(2)(vi).]

§50.73(a)(2)(v)

Licensees shall report: "Any event or

condition that alone could have prevented

the fulfillment of the safety function of

structures or systems that are needed to:

(A) Shut down the reactor and maintain it

in a safe shutdown condition;

(B) Remove residual heat;

(C) Control the release of radioactive

material; or

(D) Mitigate the consequences of an

accident."

§50.73(a)(2)(vi)

"Events covered in paragraph (a)(2)(v) of

this section may include one or more

personnel errors, equipment failures, and/or

discovery of design, analysis, fabrication,

construction, and/or procedural

inadequacies. However, individual

component failures need not be reported

pursuant to this paragraph if redundant

equipment in the same system was

operable and available to perform the

required safety function".

If not reported under §50.72(a) or (b)(1), licensees shall notify the NRC via the ENS as soon as

practical and in all cases within 4 hours of discovery of the event or condition and submit an

LER within 30 days.

Discussion

The level of judgment; fr reporting an event or condition under this criterion is a reasonable

expectation of preventing fulfillme-nt of a safety function. In the discussions which fo.llow.,many

of which are taken from the Statement of Considerations or from previous NU REG guidance, several different expressions such as "would have," "could have," "alone could have," and

Isreasonable doubt" are used to characterize this standard. In the staffs view, all of these

should be judged on the basis of a reasonable expectation of preventing fulfillment of the.safety

function.

65 NUREG-1022, Rev. I
As indicated in the Statement of Considerations, the intent of these criteria is to capture those

events where there would have been a failure of a safety system to properly complete a safety

function, regardless of when the failures were discovered or whether the system was needed at

the time.(27)

These criteria cover an event or condition where redundan! structures, components, or trains .of

a safety system could have failed to perform their intended function because of: one or more

personnel errors, including procedure violations; equipment failures; inadequate maintenance;

or design, analysis, fabrication, equipment qualification, construction, or procedural

deficiencies. The event must be reported .. g.fd.es "" f the..ituet.en e,,eediteMi.g the. aused

th., str.tur, or systems to be unavailable; and, regardless of whether or not an altemate safety

system could have been used to perform the safety function (e.g., high pressure core cooling

failed, but feed-and-bleed or low pressure core cooling were available to provide the safety

function of core cooling).

The definition of the systems included in the scope of these criteria is provided in the rules

themselves; it is not determined by the phrases "safety-related" and "important to safety."

In determining the reportability of an event or condition that affects a system, it is not necessary

to assume an additional random single failure in that system.

The term "safety function" refers to any of the four functions (A through D) listed in these

reporting criteria that are required during any plant mode or accident situation as described or

relied on in the plant safety analysis report or required by the regulations.

A system must operate long enough to complete its intended safety function as defined in the

safety analysis report. Reasonable operator actions to correct minor problems may be

considered; however, heroic actions and unusually perceptive diagnoses, particularly during

stressful situations, should not be assumed. If a potentially serious human error is made that

could have prevented fulfillment of a safety function, but recovery factors resulted in the error

being corrected, the error is still reportable.

Both offsite electrical power (transmission lines) and onsite emergency power (usually diesel

generators) are considered to be separate functions by GDC 17. If either offsite power or onsite

emergency power is unavailable to the plant (i.e. completely lost), it is reprtable regardless'of

whether the other system is available. GDC 17 defines the safety function of each system as

providing sufficient capacity and capability, etc., assuming that the other system is not

available. Loss of offsite power should be determined at the essential switchgear busses.

As indicated in the Statement of Considerations:

"The Commission recognizes that the application of this and other paragraphs of this

section involves the use of engineering judgment. In this case, a technical judgment

must be made whether a failure or operator action that did actually disable one train of a

(27) 48 FR 33854, July 28, 1983.

NUREG-1022, Rev. 1 66
safety system, could have, but did not, affect a redundant train within the ESF system.

If so, this would constitute an event that "could have prevented" the fulfillment of a

safety function, and, accordingly, must be reported.

If a component fails by an apparently random mechanism it may or may not be

reportable if the functionally redundant component could fail by the same mechanism.

Reporting is required if the failure constitutes a condition where there is reasonable

doubt that the functionally redundant train or channel would remain operational until it

completed its safety function or is repaired. For example, if a pump in one train of an

ESF system fails because of improper lubrication, and engineering judgment indicates

that there is a reasonable expectation that the functionally redundant pump in the other

train, which was also improperly lubricated, would have also failed before it completed

its safety function, then the actual failure is reportable and the potential failure of the

functionally redundant pump must be discussed in the LER.

For systems that include three or more trains, the failure of two or more trains should be

reported if, in the judgment of the licensee, the functional capability of the overall system

was jeopardized.'%28

and:

"Finally, the Commission recognizes that the licensee may also use engineering

judgment to decide when personnel actions could have prevented fulfillment of a safety

function. For example, when an individual improperly operates or maintains a

component, he might conceivably have made the same error for all of the functionally

redundant components (e.g., if he incorrectly calibrates one bistable amplifier in the

Reactor Protection System, he could conceivably incorrectly calibrate all bistable

amplifiers). However, for an event to be reportable it is necessary that the actions

actually affect or involve components in more than one train or channel of a safety

system, and the result of the actions must be undesirable from the perspective of

protecting the health and safety of the public. The components can be functionally

redundant (e.g, two pumps in different trains) or not functionally redundant (e.g., the

operator correctly stops a pump in Train "A" and instead of shutting the pump discharge

valve in Train "A," he mistakenly shuts the pump discharge valve in Train "B").'V)

Any time a system did not or could not have performed its safety function because of a single

failure, common-mode failure, or combination of independent failures it is reportable under

these criteria. These reporting requirements apply to the system level, rather than the train or

component level.

r•} 48 FR 33854 and 48 FR 33858, July 26, 1983.

(29) 48 FR 33854 and 48 FR 33858, July 26, 1983.

67 NUREG-1022, Rev. I
0 Single Failure

These reporting criteria are not meant to require reporting of a single, independent (i.e.,

random) component failure that makes only one functionally redundant train inoperative

unless it is indicative of a generic problem (i.e., has common-mode failure implications).

As indicated in Paragraph 50.73(a)(2)(vi) "...individual component failures need not be

reported pursuant to this paragraph if redundant equipment in the same system was

operable and available to perform the required safety function..."

The staff considers application of this principle to include cases where one train of a two

train system is:

(1) failed, or;

(2) othese incapable of performing its function because of factors such as

Operator believes design, analysis, fabrication, construction and/or procedural

inadequacies, or,

(3) in the oaa train rcase whichsshould be running, otherwise not performing its

function beca -use of factors such as operator error or design, analysis.,

fabrication, construction aid/or procedural inadequacies, or;

(4) otherwise subject to a reasonablaeexpectation of bing preven~ted from fulfilling

its safety function.

The staff believes that the conditions necessary to consider the redundant train operable

and available, for this purpose, should include the following:

(1) in cases where the redundant train should operate automatically, it is capable of

timely and correct automatic operation, or in cases where the redundant train

shouild be oper .ated manually, the operators would deteclP4.the need for its

operation and initiate such ope~ratio~n, .using est~ablished procedu~res.for whfich

they are trained, within the needed time frame, without the need for trouble

shooting and repair, and;

(2) the redundant train is capable of performing its safety function for the duration

required, and;

IM For example, conditions that Wou.dIndin.ca~te a need for ope~ratio~n of the.redundan~t

train are regularly monitored and instrumentation used to monitor these conditions is capable

and available.

NUREG-1022, Rev. 1 68
(3) there is not a reasonable expectation of preventing fulfillment of the safety

function by the redundant train.01)

A single failure that defeats the safety function of a system is reportable even if the

design of the system, which allows such a single failure to defeat the function of the

system, has been found acceptable.

As discussed in the Statement of Considerations, "there are a limited number of singletrain systems that perform safety functions (e.g., the High Pressure Coolant Injection

System in BWRs). For such systems, loss of the single train would prevent the

fulfillment of the safety function of that system and, therefore, is reportable even though

the plant technical specifications may allow such a condition to exist for a limited

time.'"2)

* Common-Cause Failures

The following conditions are reportable under these criteria:

an event or condition that disabled multiple trains of a system

because of a single cause

an event or condition where one train of a system is disabled; in addition, (1) the

underlying cause that disabled one train of a system could have failed a

redundant train and (2) there is reasonble xctatioh that the second train

would Mot complete its safety function if called upon

an observed or identified event or condition that alone could have prevented

fulfillment of the safety function

* Multiple equipment inoperability or unavailability

Whenever an event or condition exists where the system could have been prevented

from fulfilling its safety function because of one or more reasons for equipment

_or T example, this means th~at the exclusion from reporting single component

failures -under this criterion- (ie.-, Paragra phs 5O.72(b)(2)fffi,- 50.73(a)X2)fv) and 50OI73(a)(2)(vi~)

should not apply when there is a reasonable expectation of failure -of the redundant train as a

result of -he same cause. Application of this principle is illustrated in several parts of this

section, including: (1) the immediately proceeding quotations from 48 FR 33854 and 48 FR

33858; (2) the immediately following discussion of common cause failures, and; (3) the

discussions in Examples 12 and 13. As indicated in the first paragraph of this section, the

event should be reported under this criterion if there is a reasonable expectation of preventing

fulfillment of the safety function.

(32) 48 FR 33854, July 26, 1983.

69 NUREG-1022, Rev. 1
inoperability or unavailability, it is reportable under these criteria. This would include

cases where one train is disabled and a second train fails a surveillance test.

Reportability of any of the above type failures (single, common-mode, or multiple) under both

10 CFR 50.72 and 50.73 is independent of power or plant mode. It also is independent of

whether:

0 the system or structure was demanded at the time of discovery

* the system or structure was required to be operable at the time of discovery

0 the cause of a potential failure of the system was corrected before an actual demand for

the safety function could occur

* other systems or structures were available that could have or did perform the safety

function

0 the entire system or structure is specified as ESF or safety related

* the problem occurs in a non-safety portion of a system

The following types of events or conditions generally are not reportable under these criteria:

failures that affect inputs or services to systems that have no safety function (unless it

could prevent the performance of a safety function of an adjacent or interfacing system)

a single defective component that was delivered, but not installed

removal of a system or part of a system from service as part of a planned evolution for

maintenance or surveillance testing when done in accordance with an approved

procedure and the plant's TS (unless a condition is discovered that could have

prevented the system from performing its function)

* independent failure of a single component (unless it is indicative of a generic problem, it

alone could have caused a safety system failure, or it is in a single-train system)

a procedure error discovered before procedure approval and the error could have

resulted in defeating the system function

* a failure of a system used only to warn the operator where no credit is taken for it in any

safety analysis and it does not directly control any of the safety functions in the criteria

* a single stuck control rod that alone would not have prevented the fulfillment of a reactor

shutdown

0 unrelated component failures in several different safety systems

NUREG-1022, Rev. 1 70
The applicability of these criteria includes those safety systems designed to mitigate the

consequences of an accident (e.g., containment isolation, emergency filtration). Hence, minor

operational events involving a specific component such as valve packing leaks, which could be

considered a lack of control of radioactive material, should not be reported under this

paragraph. System leaks or other similar events may, however, be reportable under other

sections of the rules.Pt )

Examples

Single Train Systems

(1) Failure of a Single-Train System Preventing Accident Mitigation and Residual Heat

Removal

When the licensee was preparing to run a surveillance test, a high-pressure coolant

injection (HPCI) flow controller was found inoperable; therefore, the licensee declared

the HPCI system inoperable. The plant entered a technical specification requiring that

the automatic depressurization, low-pressure coolant injection, core spray, and isolation

condenser systems remain operable during the 7-day LCO or the plant had to be shut

down. The licensee made an ENS notification within 28 minutes and a followup call

after the amplifier on the HPCI flow transmitter was fixed and the HPCI retumed to

operability.

As discussed above, the loss of a single train safety system such as BWR HPCI is

reportable.

(2) Failure of a Single-Train Non-Safety System

Question:

If RCIC is not a "safety system" in that no credit for its operation is taken in the safety

analysis, are failures and unavailability of this system reportable?

Answer:

If the plant's safety analysis considered RCIC as a system needed to remove residual

heat (e.g., it is included in the Technical Specifications) then its failure is reportable

under this criterion; otherwise, it is not reportable under this section of the rule.

f) 48 FR 33854, July 26, 1983.

71 NUREG-1022, Rev. I
(3) Failure of a Single-Train Environmental System

Question:

There are a number of environmental systems in a plant dealing with such things as low

level waste (e.g., gaseous radwaste tanks). Many of these systems are not required to

meet the single failure criterion so a single failure results in the loss of function of the

system. Are all of these systems covered within the scope of the LER rule?

Answer:

If such systems are required by Technical Specifications to be operational and the

sysemisneded to fulfill one of the safety functions identified in this section of the ruts.

then system level failures are reportable. If the system is not covered by Technical

Specifications and is not required to meet the single failure criterion, then failures of the

system are not reportable under this criterion.

Loss of Two Trains

(4) Loss of Onsite Emergency Power by Multiple Equipment Inoperability and Unavailability

During refueling, one emergency diesel generator (EDG) in a two train system was out

of service for maintenance. The second EDG was declared inoperable when it failed its

surveillance test.

An ENS notification is required and an LER is required. As addressed in the Discussion

section above, loss of either the onsite power system or the offsite power system is

reportable under this criterion.

(5) Procedure Error Prevents Reactor Shutdown Function

The unit was in mode 5 (95OF and 0 psig ; before initial criticality) and a postmodification test was in progress on the train A reactor protection system (RPS), when

the operator observed that both train A and B source range detectors were disabled.

During post-modification testing on train A RPS, instrumentation personnel placed the

train B input error inhibit switch in the inhibit position. With both trains' input error inhibit

switches in the inhibit position, source range detector voltage was disabled. The input

error inhibit switch was immediately returned to the normal position and a caution was

added to appropriate plant instructions.

This event is reportable because disabling the source range detectors could have

prevented fulfillment of the safety function to shut down the reactor.

(6) Failure of the Overpressurization Mitigation System

The RCS was overpressurized on two occasions during startup following a refueling

outage because the overpressure mitigation system (OMS) failed to operate. The

NUREG-1022, Rev. 1 72
reason that the OMS failed to operate was that one train was out of service for

maintenance and a pressure transmitter was isolated and a summator failed in the

actuation circuit on the other train.

The event is reportable because the OMS failed to perform its safety function.

(7) Loss of Salt Water Cooling System and Flooding in Saltwater Pump Bay

During maintenance activities on the south saltwater pump, the licensee was removing

the pump intemals from the casing when flooding of the pump area occurred. The north

saltwater pump was secured to prevent pump damage.

The event is reportable because of the failure of the saltwater cooling system, which is

the ultimate heat sink for the facility, to perform its safety function.

(8) Maintenance Affecting Two Trains

Question:

Some clarification is needed for events or conditions that alone "could have" prevented

the fulfillment of a system safety function.

Answer:

"Events or conditions" generally involve operator actions and/or component failures that

could have prevented the functioning of a safety system. For example, assume that a

surveillance test is run on a standby pump and it seizes. The pump is disassembled

and found to contain the wrong lubricant. The redundant pump is disassembled and it

also has the same wrong lubricant. Thus, it is reasonable to assume that the second

pump would have failed if it had been challenged. However, the second pump and,

therefore, the system did not actually fail because the second pump was never

challenged. Thus, in this case, because of the use of the wrong lubricant, the system

"could have" or "would have" failed.

Loss of One Train

(9) Oversized Breaker Wiring Lugs

Situation:

During testing of 480 volt safety-related breakers, one breaker would not trip electrically.

Investigation revealed that one wire of the pigtail on the trip coil, although still in its lug,

was so loose that there was no electrical connection. The loose connection was due to

the fact that the pigtail lug was too large (No. 14-16 AWG), whereas the pigtail wire was

No. 20 AWG. A No. 18-22 lug is the acceptable industry standard for a No. 20 AWG

wire.

73 NUREG-1022, Rev. I
Since the trip coils were supplied pre-wired, all safety-related breakers utilizing the trip

coil were inspected. All other breakers inspected had No. 14-16 AWG lugs. No lugs

were found with loose electrical connections. Nevertheless, all No. 14-16 AWG lugs

were replaced with acceptable industry Standard No. 18-22 AWG lugs.

Comment:

The event is reportable because the incompatible pigtails and lugs could have caused

one or more safety systems to fail to perform their intended function [50.73(a)(2)(v)].

(10) Contaminated Hydraulic Fluid Degrades MSIV Operation

Situation:

During a routine shutdown, the operator noted that the #11 MSIV closing time appeared

to be excessive. A subsequent test revealed the #11 MSIV shut within the required

time, however, the #12 MSIV closing time exceeded the maximum at 7.4 sec.

Contamination of the hydraulic fluid in the valve actuation system had caused the

system's check valves to stick and delay the transmission of hydraulic pressure to the

actuator. Three more filters will be purchased providing supplemental filtering for each

MSIV. Finer filters will be used in pump suction filters to remove the fine contaminants.

The #12 MSIV was repaired and returned to service. Since the valves were not required

for operation at the time of discovery, the safety of the public was not affected.

Comments:

The event is reportable because a single condition could have prevented fulfillment of a

safety function [50.73(a)(2)(v)].

The fact that the condition was discovered when the valves were not required for

operation does not affect the reportability of the condition.

(11) Diesel Generator Lube Oil Fire Hazard

Situation:

While performing a routine surveillance test of the emergency diesel generator, a small

fire started due to lubricating oil leakage from the exhaust manifold. The manufacturer

reviewed the incident and determined that the oil was accumulating in the exhaust

manifold due to leakage originating from above the upper pistons of this vertically

opposed piston engine. The oil remaining above the upper pistons after shutdown

leaked slowly down past the piston rings, into the combustion space, past the lower

piston rings, through the exhaust ports, and into the exhaust manifolds. The exhaust

manifolds became pressurized during the subsequent startup which forced the oil out

through leaks in the exhaust manifold gaskets where it was ignited.

NUREG-1022, Rev. 1 74
Similar events occurred previously at this plant. In these previous cases, fuel oil

accumulated in the exhaust manifold due to extended operation under "no load"

conditions. Operation under loaded conditions was therefore required before shutdown

in order to burn off any accumulated oil.

Comments:

The event is not reportable if the fire did not pose a threat to the plant (i.e., it only

affected a single component) [50.73(a)(2)(x)].

The event would be reportable if it demonstrates a design, procedural, or equipment

deficiency that could have prevented the fulfillment of a safety function (i.e., if the

redundant diesels are of similar design and, therefore, susceptible to the same problem)

[50.73(a)(2)(v)].

(12) Single Failures

Question:

I notice that loss of relief/safety valve capability is reportable. Does this mean that an

LER is required when one valve is inoperative? In addition, suppose your have one

pump in a cooling water system (e.g., chilled water) supplying water to both trains of a

safety system, but there is another pump in standby; is the loss of the one operating

pump reportable?

Answer:

No. Single, independent (i.e., random) component failures are not reportable as LERs if

the redundant component in the same system did or would have fulfilled the safety

function. !,n ge'neraltw s"er, •6ai"uro--arc * * Fr 60•rb1c 46th& NPRD ý-ys-'ft i

However, if such failures have generic implications, then an LER is to be submitted.

(See the discussion under the heading "Single Failures" for further discussion of

reporting the loss of one train.)

(13) Generic Setpoint Drift

0 Situation:

With the plant in steady state operation at 2170 MWt and while performing a

Main Steam Line Pressure Instrument Functional Test and Calibration, a switch

was found to actuate at 853 psig. The Tech Specs limit is 825 +15 psig. The

redundant switches were operable. The cause of the occurrence was setpoint

drift. The switch was recalibrated and tested successfully per HNP-2-5279,

Barksdale Pressure Switch Calibration, and returned to service.

75 NUREG-1022, Rev. 1
This is a repetitive event as reported in one previous LER. A generic review

revealed that these type switches are used on other safety systems and that this

type switch is subject to drift. An investigation will continue as to why these

switches drift, and if necessary, they will be replaced.

Comments:

The event is not reportable due to the drift of a single pressure switch.

The event is reportable if it is indicative of a generic and/or repetitive problem

with this type of switch which is used in several safety systems [50.73(a)(2)(v) or

(viii)].

* Question:

Are setpoint drift problems with a particular switch to be reported if they are

experienced more than once?

Answer:

The independent failure (e.g., excessive setpoint drift) of a single pressure switch

is not reportable unless it alone could have caused a system to fail to fulfill its

safety function, or is indicative of a generic problem that could have resulted in

the failure of more than one switch and thereby cause one or more systems to

fail to fulfill their safety function.

(14) Maintenance Affecting Only One Train

Question:

Suppose the wrong lubricant was installed in one pump, but the pump in the other train

was correctly lubricated. Is this reportable?

Answer:

Engineering judgement is required to decide if the lubricant could have been used on

the other pump, and, therefore, the system function would have been lost. If the

procedure called for testing of the first pump before maintenance was performed on the

second pump and testing clearly identified the error, then the error would not be

reportable. However, if the procedure called for the wrong lubricant and eventually both

pumps would have been improperly lubricated, and the problem was only discovered

when the first pump was actually challenged and failed, then the error would be

reportable.

NUREG-1022, Rev. 1 76
Other Conditions

(15) Conditions Observed While System Out of Service

Question:

Suppose during shutdown we are doing maintenance on both SI pumps, which are not

required to be operational. Is this reportable? While shutdown, suppose I identify or

observe something that would cause the SI pumps not to be operational at power. Is

this reportable?

Answer:

Removing both SI pumps from service to do maintenance is not reportable if the

resulting system configuration is not prohibited by the plant's technical specifications.

However, if a situation is discovered during maintenance that could have caused both

pumps to fail, (e.g., they are both improperly lubricated) then that condition is reportable

even though the pumps were not required to be operational at the time that the condition

was discovered. As another example, suppose the scram breakers were tested during

shutdown conditions, and it was found that for more than one breaker, opening times

were in excess of those specified, or that UV trip attachments were inoperative. Such

potential generic problems are reportable in an LER.

(16) Diesel Generator Bearing Problems

During the annual inspection of one standby diesel generator, the lower crankshaft

thrust bearing and adjacent main bearing were found wiped on the journal surface. The

thrust bearing was also found to have a small crack from the main oil supply line across

the journal surface to the thrust surface. Inspection of the second, redundant standby

diesel generator annual inspection revealed similar problems. It was judged that

extended operation without corrective action could have resulted in bearing failure.

The event is reportable because there was reasonable doubt that both diesels would

have remained operable until they completed their safety function if called upon.

(17) Potential Loss of High Pressure Coolant Injection

During normal refueling leak testing of the upstream containment isolation check valve

on the High Pressure Coolant Injection (HPCI) steam exhaust, the disc of the noncontainment isolation check valve was found lodged in downstream piping. This might

have prevented HPCI from functioning if the disc had blocked the line. The event was

caused by fatigue failure of a disc pin.

Following evaluation of the condition, the event was determined to be reportable

because the HPCI could have been prevented from performing its safety function if the

disc had blocked the line. In addition, the event is reportable if the fatigue .failure is

indicative of a common-mode failure.

77 NUREG-1022, Rev. 1
(18) Defective Component Delivered but not Installed

Question:

How should a plant report a defective component that was delivered, but not installed?

Answer:

A single defective component would not generally be reportable (assuming that the

problem has no generic implications). A generic problem or a number of defective

components would probably constitute a condition that could have prevented fulfillment

of a safety function, and, if so, would 'e reportable. Engineering judgment is required to

determine if the defects could have tscaped detection prior to installation and operation.

As a minimum, any generic problem may be reported as a voluntary LER. In addition,

such a condition may be reportable under 10 CFR Part 21.

(19) Operator Inaction or Wrong Action

Question:

In some systems used to control the release of radioactivity, a detector controls certain

equipment. In other systems, a monitor is present and the operator is required to initiate

action under certain conditions. The operator is not 'wired" in. Are failures of the

operator to act reportable?

Answer:

Yes. The operator may be viewed as a "component" that is an integral, and frequently

essential, part of a "system." Thus, if an event or condition meets the criterion specified

in 50.73 for reporting, it is to be reported regardless of the initiating cause (i.e., whether

an equipment, procedure, or personnel error is involved).

(20) Results of Analysis

Question:

A number of criteria indicate that they apply to actual situations only and not to potential

situations identified as a result of analysis; yet, other criteria address "could have."

When do the results of analysis have to be reported?

Answer

The results need only to be reported if the applicable criterion requires the reporting of

conditions that "could have" caused a problem. However, others have a need to know

about potential problems that are not reportable; thus, such items may be reported as a

voluntary LER.

NUREG-1022, Rev. I 78
(21) System Interactions

Question:

Utilities are not required to analyze for system interactions, yet the rule requires the

reporting of events that "could have" happened but did not. Are we to initiate a design

activity to determine "could have" system interactions?

Answer:

No. Report system interactions that you find as a result of ongoing routine activities

(e.g., the analysis of operating events).

79 NUREG-1022, Rev. 1
3.3.4 Common-cause Failures of Independent Trains or Channels

*1

10 CFR 50.72 §50.73(a)(2)(vi0)

[No corresponding Part 50.72 requirement.] Licensees shall report: "Any event where a

single cause or condition caused at least

one independent train or channel to become

inoperable in multiple systems or two

independent trains or channels to become

inoperable in a single system designed to:

(A) Shut down the reactor and maintain it in

a safe shutdown condition;

(B) Remove residual heat;

(C) Control the release of radioactive

material; or

(D) Mitigate the consequences of an

accident."

Licensees are required to report a common-cause failure as an LER within 30 days.

Discussion

This criterion requires those events to be reported where a single cause or condition caused

independent trains or channels to become inoperable. Common-causes may includ suoh.

facton as high amthbent te npoetures, heatup from aiergization, inadequate preventive

maintenance; oil contain a tai of air systems, incorrect lubrication, use of nonqule if [e

compnonents or manufact ring or desin flaws. The event is reportable if indepe1dent trhn ,te 

or channels were noperablegat the same time, regardless of whether or not they were

discovered at the same time.- (Example (2) below illustrates a- cas where the second failure

was discovered 3 days later than the first.)

An event or failure that results in or involves the failure of independent portions of more than

one train or channel in the same or different systems is reportable. For example, if a cause or

condition caused components in Train "A" and "B" of a single system to become inoperable,

even if additional trains (e.g., Train "C") were still available, the event must be reported. In

addition, if the cause or condition caused components in Train "A" of one system and in Train

"B" of another system (iLe., train that is assumed in the safety analysis to be independent) to

become inoperable, the event must be reported. However, if a cause or condition caused

components in Train "A" of one system and Train "A" of another system (i.e., trains that are not

assumed in the safety analysis to be independent), the event need not be reported unless it

meets one or more of the other reporting criteria.

Trains o*:r channels for repcnrtability purposes are definhed as those re'dun~dant; independentj trains or channels designed to provide protection against single fail!Ures,. Many engineere safety systems containing-. actWe components are designed with at least a two-train systemNUREG-1022, Rev. 1 80
Each independent train in a two-train systemn can normally satisfy all the safety system

requirements to safely shut down the plant or satisfy those criteria that have to be met following

an accident.

This criterion does not include those cases where one train of a system or a component was

removed from service as part of a planned evolution, in accordance with an approved

procedure, and in accordance with the plant's technical specifications. For example, if the

licensee removes part of a system from service to perform maintenance, and the Technical

Specifications permit the resulting configuration, and the system or component is returned to

service within the time limit specified in the Technical Specifications, the action need not be

reported under this paragraph. However, if, while the train or component is out of service, the

licensee identifies a condition that could have prevented the whole system from performing its

intended function (e.g, the licensee finds a set of relays that is wired incorrectly), that condition

must be reported.

Analysis of events reported under this part of the rule may identify previously unrecognized

common-cause (or dependent) failures and system interactions. Such failures can be

simultaneous failures that occur because of a single initiating cause (i.e., the single cause or

mechanism serves as a common input to the failures); or the failures can be sequential (i.e.,

cascading failures), such as the case where a single component failure results in the failure of

one or more additional components.

Examples

(1) Incorrect Lubrication Degrades Main Steam Isolation Valve Operation

During monthly operability tests, the licensee found that the Unit 2B inboard MSIV did

not stroke properly as a result of a solenoid-operated valve (SOV) failure. Both units

were shut down from 100-percent power, and the SOVs piloting all 16 MSIVs were

inspected. The licensee found that the SOVs on all 16 MSIVs were damaged. The

three-way and four-way valves and solenoid pilot valves on all 16 MSIVs had a

hardened, sticky substance in their ports and on their O-rings. As a result, motion of all

the SOVs was impaired, resulting in instrument air leakage and the inability to operate

all of the MSIVs satisfactorily. The licensee also examined unused spares in the

warehouse and found that the lubricant had dried out in those valves, leaving a residue.

Several of the warehouse spares were bench tested. They were found to be degraded

and also leaked. The root cause of the event was use of an incorrect lubricant.

The event is reportable (a) because a single cause or condition caused multiple

independent trains of the main steam isolation system (a system designed to control the

release of radioactive material and mitigate the consequences of an accident) to

become inoperable [§50.73(a)(2)(vii)(C and D)] and (b) because a single condition could

have prevented fulfillment of a safety function [§50.73(a)(2)(v)].

81 NUREG-1022, Rev. 1
(2) Marine Growth Causing Emergency Service Water To Become Inoperable (CommonMode Failure Mechanism)

With Unit 1 at 74 percent power and Unit 2 at 100 percent power, ESW pump 1A was

declared inoperable because its flow rate was too low to meet acceptance criteria.

Three days later, with both units at the same conditions, ESW pump 1C was declared

inoperable for the same reason. The ESW pumps provide the source of water to the

intake canal during a design-basis accident. In both cases, the cause was marine

growth of hydroids and barnacles on the impeller and suction of the pumps. Following

maintenance, both pumps passed their performance tests and were placed in service.

Pump testing frequency was increased to more closely monitor pump performance.

This event is reportable because a single cause or condition caused two independent

trains to become inoperable in a single system designed to mitigate the consequences

of an accident [§50.73(a)(2)(vii)(D)].

(3) Testing Indicated Several Inoperable Snubbers

The licensee found 11 inoperable snubbers during periodic testing. All the snubbers

failed to lock up in tension and/or compression. These failures did not render their

respective systems inoperable, but rendered trains inoperable. Improper lockup settings

and/or excessive seal bypass caused these snubbers to malfunction. These snubbers

were designed for low probability seismic events. Numerous previous similar events

have been reported by this licensee.

This condition is reportable because the condition indicated a generic common-mode

problem that caused numerous multiple independent trains in one or more safety

systems to become inoperable. The potential existed for numerous snubbers in several

systems to fail following a seismic event rendering several trains inoperable. [§

50.73(a)(2)(vii)]

(4) Stuck High-Pressure Injection (HPI) System Check Valves as a Result of Corroded

Flappers

The licensee reported that check valves in three of four HPI lines were stuck closed.

The unit had been shut down for refueling and maintenance.

A special test of the check valves revealed that three 21⁄2-inch stop check valves

remained closed when 130 pounds per square inch (psi) of differential pressure was

applied to the valve. An additional test revealed that the valve failed to open when 400

psi of differential pressure (the capacity of the pump) was applied to the valve. Further

review showed that the common cause of valve failure was the flappers corroding shut.

The event is reportable because a single cause or condition caused at least two

independent trains of the HPI system to become inoperable. This system is designed to

remove residual heat and mitigate the consequences of an accident. The condition is

NUREG-1022, Rev. 1 82
therefore reportable under 50.73(a)(2)(vii)(B and D), common cause failure in systems

designed to remove residual heat and mitigate accidents.

83 NUREG-1022, Rev. 1
3.3.5 Airborne or Liquid Effluent Release

§50.72(b)(2)(iv) §50.73(a)(2)(viii)

Licensees shall report: Licensees shall report:

"(A) Any airborne radioactive release that, "(A) Any airborne radioactivity release that,

when averaged over a time period of 1 hour, when averaged over a time period of 1 hour,

results in concentrations in an unrestricted resulted in airborne radionuclide

area that exceed 20 times the applicable concentrations in an unrestricted area that

concentration specified in Appendix B to exceeded 20 times the applicable

Part 20, Table 2, Column 1. concentration limits specified in Appendix B

(B) Any liquid effluent release that, when to Part 20, Table 2, Column 1.

averaged over a time period of 1 hour, (B) Any liquid effluent release that, when

exceeds 20 times the applicable averaged over a time period of 1 hour,

concentration specified in Appendix B to exceeds 20 times the applicable

Part 20, Table 2, Column 2, at the point of concentrations specified in Appendix B to

entry into the receiving waters (i.e., Part 20, Table 2, Column 2, at the point of

unrestricted area) for all radionuclides entry into the receiving waters (i.e.,

except tritium and dissolved noble gases. unrestricted area) for all radionuclides

(Immediate notifications made under this except tritium and dissolved noble gases.

paragraph also satisfy the requirements of

§20.2202 of this chapter.)" §50.73(a)(2)(ix)

Reports submitted to the Commission in

accordance with paragraph (a)(2)(viii) of this

section also meet the effluent release

reporting requirements of §20.2203(a)(3) of

If not reported under §50.72(a) or (b)(1), licensees are required to report such airborne or liquid

effluent releases as defined in the regulations above to the NRC via the ENS as soon as

practical and in all cases within 4 hours of the event. Licensees are required to submit an LER

within 30 days.

Discussion

Although similar to 10 CFR 20.403 (20.2202) and 20.405 (20.2203), these criteria place a lower

threshold for reporting events at commercial power reactors because the significance of the

breakdown of the licensee's program that allowed such a release is the primary concern, rather

than the significance of the effect of the actual release.

For a release that takes less than 1 hour, normalize the release to 1 hour (e.g., if the release

lasted 15 minutes, divide by 4). For releases that lasted more than 1 hour, use the highest

release for any continuous 60-minute period (i.e., comparable to a moving average).

NUREG-1022, Rev. 1 84
Annual average meteorological data should be used for determining offsite airborne

concentrations of radioactivity to maintain consistency with the technical specifications (TS) for

reportability thresholds.

The location used as the point of release for calculation purposes should be determined using

the expanded definition of an unrestricted area as specified in NUREG-0133 ("Preparation of

Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978) to

maintain consistency with the TS.

If estimates determnIne that the rolease has exceeded the reporting criterion, ;an ENS notificaition

Is required, followed up by a more Iprecise e .stimate in the LER. If it is later dIetermined that the,

release was less than, this criterion, the ENS notification may-be retracted,

As indicated in Generic Letter 85-19, September 27, 1985, "Reporting Requirements on

Primary Coolant Iodine Spikes," primary coolant iodine spike releases need not be reported on

a short term basis.

Examp-ies

(1) Unmonitored Release of Contaminated Steam Through Auxiliary Boiler Atmospheric

Vent

An unmonitored release of contaminated steam resulted from a combination of a tube

leak, improper venting of an auxiliary boiler system, and inadequate procedures. This

combination resulted in a release path from a liquid waste concentrator to the

atmosphere via the auxiliary boiler system steam drum vent.

Because of rain at the site, the steam release to the atmosphere was condensed and

deposited onto plant buildings and yard areas. This contamination was washed via a

storm drain into a lake. The release was later confirmed to be 2.6 E-5 pCi/mI of Cs-137

at the point of entry into the receiving water.

An ENS notification is required as a liquid radioactive material release because the

unmonitored release exceeded 20 times the applicable concentrations specified in Table

2, Column 2 of Appendix B to 10 CFR Part 20, averaged over 1 hour at the site

boundary. An LER is required.

(2) Unplanned Gaseous Release

During routine scheduled maintenance on a pressure actuated valve in the gaseous

waste system, an unplanned radioactive release to the environment was detected by a

main stack high radiation alarm. The release occurred when an isolation valve, required

to be closed on the station tagout sheet, was inadvertently left open. This allowed

radioactive gas from the waste gas decay tank to escape through a pressure gage

connection that had been opened to vent the system. Operator error was the root

cause of this release, with ambiguous valve tag numbers as a contributing factor. The

85 NUREG-1022, Rev. 1
concentration in the unrestricted area, averaged over I hour, was estimated by the

licensee to be 1 E-5 pCVml of Kr-85 and 5 E-6 pCi/ml of Xe-1 33.

The event was reportable via ENS and LER because the sum of the ratios of the

concentration of each airborne radionuclide in the restricted area when averaged over a

period of 1 hour, to its respective concentration specified in Table 2, Column 1 of

Appendix B to 10 CFR 20, exceeds 20.

NUREG-1022, Rev. 1 86
3.3.6 Contaminated Person Requiring Transport Offsite

§50.72(b)(2)(v) 10 CFR 60.73

Licensees shall report: "Any event requiring [No corresponding Part 50.73 requirement.]

the transport of a radioactively contaminated

person to an offsite medical facility for

treatment."

If not reported under §50.72(a) or (b)(1), licensees are required to notify the NRC via the ENS

of any such transport as soon as practical and in all cases within 4 hours of the event

necessitating the offsite transport.

Discussion

The phrase "radioactively contaminated" refers to either radioactively contaminated clothing

and/or person. Iffthere is aPhti1W•ft•i contamintito e(b.b.h ifitial ontitte Buryeyfot

adotive contamination Is-requrdbihsno enonpte before transport of the person

Off sitefor m~edical treatment) the licensee should make an ENS notification. See the exam p-l-e-.-

No LER is required for transporting a radioactively contaminated person to an offsite medical

facility for treatment.

Example

(1) Radioactively Contaminated Person Transported Offsite for Medical Treatment

A contract worker experienced a back injury lifting a tool while working in the reactor

containment and was considered potentially contaminated because his back could not

be surveyed. Health physics (HP) technicians accompanied the worker to the hospital.

The licensee made an ENS notification immediately and an update notification after

clothing, but not the individual, was found to be contaminated. The HP technicians

returned to the plant with the contaminated protective clothing worn by the worker.

If not reported under §50.72(a)(1) as a declared Unusual Event per the licensee's

emergency plan, an ENS notification is required because of the transport of a

radioactively contaminated person to an offsite medical facility for treatment.

87 NUREG-1022, Rev. 1
3.3.7 News Release or Other Government Notifications

§50.72(b)(2)(vI) 10 CFR 50.73

Licensees shall report: "Any event or [No corresponding Part 50.73 requirement.]

situation, related to the health and safety of

the public or on-site personnel, or protection

of the environment, for which a news

release is planned or notification to other

government agencies has been or will be

made. Such an event may include an onsite fatality or inadvertent release of

radioactively contaminated materials."

If not reported under §50.72(a) or (b)(1), licensees are required to notify the NRC via the ENS

as soon as practical and in all cases within 4 hours of the event, or the decision to prepare a

news release, or the decision to notify (or actual notification of) other government agencies.

Discussion

The purpose of this criterion is to ensure the NRC is made aware of issues that will cause

heightened public or government concern related to the radiological health and safety of the

public or on-site personnel or protection of the environment.

Licensees typically issue press releases or notify local, county, State or Federal agencies on a

wide range of topics that are of interest to the general public. The NRC Operations Center

does not need to be made aware of every press release made by a licensee. The following

clarifications are intended to set a reporting threshold that ensures necessary reporting, while

minimizing unnecessary reporting.

Examples of events likely to be reportable under this criterion include

* release of radioactively contaminated tools or equipment to public areas

* unusual or abnormal releases of radioactive effluents

* onsite fatality

Licensees generally do not have to report media and government interactions unless they are

related to the radiological health and safety of the public or onsite personnel, or protection of

the environment. For example, the NRC does gealneed to be inflrmedu•nuderthi

criterion of..

A mi.rdeViationt from sewage..or chlorine effluient limits

* minor non-radioactive, onsite chemical spills~i

*problems with plant stack or water tower aviatibn light riI

NUREG-1022, Rev. 1 88
* Peaceful d#,monsWtiatkrns

* routine reports of effluent releases to other agencies

Press Release

The NRC has an obligation to inform the public about issues within the NRC's purview that

affect or raise a concern about the public health and safety. Thus, the NRC needs accurate,

detailed information in a timely manner regarding such situations. The NRC should be aware of

information that is available for the press or other government agencies.

However, the NRC need not be notified of every press release a licensee issues. The field of

NRC interest is narrowed by the phrase "related to the health and safety of the public or onsite

personnel, or protection of the environment," in order to exclude administrative matters or those

events of no safety significance.

Routine radiation releases are not specifically reportable under this criterion. However, if a

release receives media attention, the release is reportable under this criterion.

If possible, licensees should make an ENS notification before issuing a press release because

news media representatives will usually contact the NRC public affairs officer shortly after its

issuance for verification, explanation, or interpretation of the facts.

Other Government Notifications

For reporting purposes, "other government agencies" refers to local, State or other Federal

agencies.

Notifying another Federal agency does not relieve the licensee of the requirement to report to

the NRC.

For those plants which provide a State incident response facility with alarm indication coincident

with control room alarms, e.g., an effluent radiation monitor alarm, but the actual radiation

release is less than the criteria in §50.72(b)(2)(iv), the NRC does not consider these alarm

indications as a notification to the State by the licensee. An alarm received at a State facility is

in itself not a requirement for notifying the NRC. In so far as this reporting criterion is

concerned, the licensee need only notify the NRC when the licensee determines that a

reportable release has occurred, or believes a real potential exists for interest on the part of the

State, the media, or the public, or a press release is being planned.

Examples

(1) Onsite Drowning Government Notifications and Press Release

A boy fell into the discharge canal while fishing and failed to resurface. The licensee

notified the local sheriff, State Police, U.S. Coast Guard and State emergency agencies.

Local news agencies were granted onsite access for coverage of the event. The

licensee notified the NRC resident inspector.

89 NUREG-1022, Rev. 1
As ENS notification is needed because of the fatality on-site, the other government

notifications made, and media involvement.

(2) Licensee Media Inquiries Regarding NRC Findings

As a result of a local newspaper article regarding the findings of an NRC regional

inspection of the 10 CFR Part 50, Appendix R, Fire Protection Program, a licensee

representative was interviewed on local television and radio stations. The licensee

notified State officials and the NRC resident inspector.

The* staff doads ho t- con-hside r a n ENS notif icatio n to be hne*eded b e-a u- -e- the-su b'ect -f thr

radio and TV interviews was -an NRC inspection.

(3) County Government Notification

The licensee informed county governments and other organizations of a spurious

actuation of several emergency response sirens in a county (for about 5 minutes

according to county residents). The licensee also planned to issue a press release.

An ENS notification is needed because county agencies were notified regarding the

inadvertent actuation of part of the public notification system. Such an event also would

be reportable if the county informs the licensee of the problem because of the concern

of the public for their radiological health and safety.

(4) State Notification of Unscheduled Radiation Release

The licensee reported to the State that they were going to release about 50 curies of

gaseous radioactivity to the atmosphere while filling and venting the pressurizer. The

licensee then revised their estimate of the release to 153 curies. However, since the

licensee had not informed the State within 24 hours of making the release, they had to

reclassify the release as "unscheduled" per their agreement with the State. The

licensee notified the State and the NRC resident inspector.

An ENS notification is needed because of the State notification of an "unscheduled"

release of gaseous radioactivity. The initial notification to the State of the scheduled

release does not need an ENS notification because it is considered as a routine

notification.

(5) State Notification of Improper Dumping of Radioactive Waste

The licensee transported two secondary side filters to the city dump as nonradioactive

waste but later determined they were radioactive. The dump site was closed and the

filters retrieved. The licensee notified the appropriate State agency and the NRC

resident inspector.

NUREG-1022, Rev. 1 90
An ENS notification is needed because of the notification to the State agency of the

inadvertent release of radioactively contaminated material off site, which affects the

radiological health and safety of the public and environment.

(6) Reports Regarding Endangered Species

The licensee notified the U.S. Fish & Wildlife Service and a State agency that an

endangered species of sea turtle was found in their circulating water structure trash bar.

No press release was issued.

An ENS notification i's req'_*ured becau'se_ of the notification of State a~nd federal agencies .rgarding the taking of anijend-angered species-, (The-NRC has statutory respons-ib _ilipties

Mearciing protectionl of endangered species..)

(7) Routine Agency Notifications

A licensee notified the U.S. Environmental Protection Agency (EPA) that the circulation

water temperature rise exceeded the release permit allowable. This event was caused

by the unexpected loss of a circulating water pump while operating at 92-percent power.

The licensee reduced power to 73 percent so that the circulating water temperature

would decrease to within the allowable limits until the pump could be repaired.

A licensee notified the Federal Aviation Agency that it removed part of its auxiliary boiler

stack aviation lighting from service to replace a faulty relay.

A licensee notified the State, EPA, U.S. Coast Guard and Department of Transportation

that 5 gallons of diesel fuel oil had spilled onto gravel-covered ground inside the

protected area. The spill was cleaned up by removing the gravel and dirt.

Th'e staff does nol dohtsder gn ENS' nhotficaftion- to be56 h fleeueth§ V nts-r

routine and have little significance.debcastheevtsTr

91 NUREG-1022, Rev. I
3.3.8 Spent Fuel Storage Cask Notifications

§ 50.72(b) (2) (vii)

Licensees shall report: "Any instance of:

(A) A defect in any spent fuel storage

cask structure, system, or component which

is important to safety; or

(B) A significant reduction in the

effectiveness of any spent fuel storage cask

confinement system during use of the

storage cask under a general license issued

under §72.210 of this chapter.

A followup written report is required by

§72.216(b) of this chapter including a

description of the means employed to repair

any defects or damage and prevent

recurrence, using instructions in §72.4,

within 30 days of the report submitted in

paragraph (a). A copy of the written report

must be sent to the administrator of the

appropriate Nuclear Regulatory Commission

regional office shown in Appendix D to part

20 of this chapter."

10 CFR 50.73

[No corresponding Part 50.73 requirement.]

If not reported under §50.72(a) or (b)(1), licensees are required to report any such instances to

the NRC via the ENS as, soon as practical, and in all cases within 4 hours. A followup written

report is required by §72.216(b) within 30 days.

Discussion

This information is necessary to inform the NRC of potential hazards to the public health and

safety. The definition of "defect" in 10 CFR 21.3 is compatible with the intent of this reporting

requirement. If the defect is evaluated and reported via this reporting criterion of §50.72, then

as indicated in §21.2(c), the evaluation and notification obligations of 10 CFR Part 21 are met.

(See Section 5.1.9 for further discussion of Part 21 reporting.)

NUREG-1022, Rev. 1 92
3.4 Followup Notification

This section addresses §50.72(c), "Followup Notification." These notifications are in addition to

making the required initial telephone notifications under §50.72(a) or (b). Reporting under this

paragraph is intended to provide the NRC with timely notification when an event becomes more

serious or additional information or new analysis clarify an event. The paragraph also

authorizes the NRC to maintain a continuous communications channel for acquiring necessary

followup information.

§50.72(c)

"Followup Notification. With respect to the

telephone notifications made under paragraphs (a)

and (b) of this section, in addition to making the

required initial notification, each licensee shall,

during the course of the event:

(1) Immediately report

(i) any further degradation in the level of safety

of the plant or other worsening plant conditions,

including those that require the declaration of any

of the Emergency Classes, if such a declaration

has not been previously made, or

(ii) any change from one Emergency Class to

another, or

(iii) a termination of the Emergency Class.

(2) Immediately report

(i) the results of ensuing evaluations or

assessments of plant conditions,

(ii) the effectiveness of response or protective

measures taken, and

(iii) information related to plant behavior that is

not understood.

(3) Maintain an open, continuous communication

channel with the NRC Operations Center upon

request by the NRC."

10 CFR 60.73

[No corresponding Part 50.73

requirement.]

Discussion

These criteria are intended to provide the NRC with timely notification when an event becomes

more serious or additional information or new analyses clarify an event. They also permit the

NRC to maintain a continuous communications channel because of the need for continuing

followup information or because of telecommunications problems.

With regard to the open, continuous communications channel, licensees have a responsibility to

provide enough on-shift personnel, knowledgeable about plant operations and emergency plan

implementation, to enable timely, accurate, and reliable reporting of operating events without

93 NUREG-1022, Rev. 1
interfering with plant operation as discussed in the Statement of Considerations for the rule and

Information Notice 85-80, 'Timely Declaration of an Emergency Class, Implementation of an

Emergency Plan, and Emergency Notifications."

NUREG-1022, Rev. 1 94
4 EMERGENCY NOTIFICATION SYSTEM REPORTING

This section describes the ENS referenced in 10 CFR 50.72 and provides general and specific

guidelines for ENS reporting.

4.1 Emergency Notification System

The NRC Operations Center is the nucleus of the ENS and has the capability to handle

emergency communication needs. The NRC's response to both emergencies and nonemergencies is coordinated in this communication center. The key NRC emergency

communications personnel, the emergency officer (EO), regional duty officer (RDO), and the

headquarters operations officer (HOO), are trained to notify appropriate NRC personnel and to

focus appropriate NRC management attention on any significant event.

(1) ENS Telephones

Each commercial nuclear power reactor facility has ENS telephones funded by the NRC.

These telephones are located in each licensee's control room, technical support center

(TSC), and emergency operations facility (EOF). A separate ENS line is installed at

EOF's which are not onsite. The ENS is part of the Federal Telecommunications

System (FTS). This FTS ENS replaces the dedicated ENS ringdown telephones used

previously to provide a reliable communications pathway for event reporting.

(2) Health Physics Network Telephones

The health physics network (HPN) is designed to provide health physics and

environmental information to the NRC Operations Center in the event of an ongoing

emergency.

These telephones are installed in each licensee's TSC and EOF and, like the ENS, they

are now part of the FTS.

(3) Tape Recording

The NRC tape-records all conversations with the NRC Operations Center. The tape is

saved for a month in case there is a public or private inquiry.

(4) Facsimile Transmission (Fax)

Licensees occasionally fax an event notification into the NRC Operations Center on a

commercial telephone line before making an ENS notification. However, §50.72

95 NUREG-1022, Rev. I
requires that licensees notify the NRC Operations Center via the ENS; therefore,

licensees also must make an ENS notification.

4.2 General ENS Notification

4.2.1 Timeliness

The required timing for ENS reporting is spelled out in §§50.72(a)(3), (b)(1), (b)(2), (c)(1), (c)(2),

and in the Statements of Considerations, as 'lmmediate" and "as soon as practical and in all

cases within one (or four) hour(s)" of the occurrence of an event (depending on its significance).

The intent is to require licensees to make and act on reportability decisions in a timely manner

so that ENS notifications are made to the NRC as soon as practical, keeping in mind the safety

of the plant. See Section 2.11 for further discussion of reporting timeliness.

4.2.2 Voluntary Notifications

Licensees may make voluntary or courtesy ENS notifications about events or conditions in

which the NRC may be interested. The NRC responds to any voluntary notification of an event

or condition as its safety significance warrants, regardless of the licensee's classification of the

reporting requirement. If it is determined later that the event is reportable, the licensee can

change the ENS notification to a required notification under the appropriate 10 CFR 50.72

reporting criterion.

4.2.3 ENS Notification Retraction

If a licensee makes a 10 CFR 50.72 ENS notification and later determines that the event or

condition was not reportable, the licensee should call the NRC Operations Center on the ENS

telephone to retract the notification and explain the rationale for that decision. There is no set

time limit for ENS telephone retractions. However, since most retractions occur following

completion of engineering and/or management review, it is expected that retractions would

occur shortly after such review. See section 2.10 for further discussion of retractions.

4.2.4 ENS Event Notification Worksheet (NRC Form 361)

The ENS Event Notification Worksheet (NRC Form 361) is an attachment to Information Notice

89-89, dated December 26, 1989, subject: Event Notification Worksheets. The worksheet

provides the usual order of questions and discussion for easier communication and its use often

enables a licensee to prepare answers for a more clear and complete notification. A clear ENS

notification helps the HOO to understand the safety significance of the event.

Licensees may obtain an event number and notification time from the HOO when the ENS

notification is made. If an LER is required, the licensee may include this information in the LER

to provide a cross reference to the ENS notification, making the event easier to trace.

Licensees should use proper names for systems and components, as well as their

alphanumeric identifications during ENS notifications. Licensees should avoid using local

jargon for plant components, areas, operations, and the like so that the HOO can quickly

NUREG-1022, Rev. 1 96
understand the situation and have fewer questions. In addition, others not familiar with the

plant can more readily understand the situation.

4.3 Typical ENS Reporting Issues

At the time of an ENS notification, the NRC must independently assess the status of the reactor

to determine if it is in a safe condition and expected to remain so. The HOO needs to

understand the safety significance of each event to brief NRC management or initiate an NRC

response. The HOO will be primarily concerned about the safety significance of the event, the

current condition of the plant, and the possible near-term effects the event could have on plant

safety. The HOO will attempt to obtain as complete a description as is available at the time of

the notification of the event or condition, its causes, and its effects. Depending upon the

licensee's description of the event, the HOO may be concerned about other related issues. The

questions that the licensees typically may be asked to discuss do not represent a requirement

for reporting. These questions are of a nature to allow the HOO information to more fully

understand the event and its safety significance and are not meant in any way to distract the

licensee from more important issues.

The licensee's first responsibility during a transient is to stabilize the plant and keep it safe.

However, licensees should not delay declaring an emergency class when conditions warrant

because delaying the declaration can defeat the appropriate response to an emergency.

Because of the safety significance of a declared emergency, time is of the essence. The NRC

needs to become aware of the situation as soon as practical to activate the NRC Operations

Center and the appropriate NRC regional incident response center, as necessary, and to notify

other Federal agencies.

The effectiveness of the NRC response during an event depends largely on complete and

accurate reporting from the licensee. During an emergency, the appropriate regional incident

response center and the NRC Operations Center become focal points for NRC action.

Licensee actions during an emergency are monitored by the NRC to ensure that appropriate

action is being taken to protect the health and safety of the public. When required, the NRC

supports the licensee with technical analysis and coordinates logistics support. The NRC keeps

other Federal agencies informed of the status of an incident and provides information to the

media. In addition, the NRC assesses and, if necessary, confirms the appropriateness of

actions recommended by the licensee to local and State authorities.

Information Notice 85-80, "Timely Declaration of an Emergency Class, Implementation of an

Emergency Plan, and Emergency Notification," dated October 15, 1985, indicates that it is the

licensee's responsibility to ensure that adequate personnel, knowledge about plant conditions

and emergency plan implementing procedures, are available on shift to assist the shift

supervisor to classify an emergency and activate the emergency plan, including making

appropriate notifications, without interfering with plant operation. When 10 CFR 50.72 was

published, the NRC made clear its intent in the Statements of Consideration that notifications

on the ENS to the NRC Operations Center should be made by those knowledgeable of the

event. If the description of any emergency is to be sufficiently accurate and timely to meet the

intent of the NRC's regulations, the personnel responsible for notification must be properly

trained and sufficiently knowledgeable of the event to report it correctly. The NRC did not

97 NUREG-1022, Rev. 1
intend that notifications made pursuant to 10 CFR 50.72 would be made by those who did not

understand the event that they are reporting.

ENS reportability evaluations should be concluded and the ENS notification made as soon as

practical and in all cases within 1 hour or 4 hours to meet 10 CFR 50.72. The Statement of

Considerations noted that the 1-hour deadline is necessary if the NRC is to fulfill its

responsibilities during and following the most serious events occurring at operating nuclear

power plants without interfering with the operator's ability to deal with an accident or transient in

the first few critical minutes (48 FR 39041, August 29, 1983).

NUREG-1022, Rev. 1 98
5 LICENSEE EVENT REPORTS

This section discusses the guidelines for preparing and submitting LERs. Section 5.1

addresses administrative requirements and provides guidelines for submittal; Section 5.2

addresses the requirements and guidelines for the LER content. Portions of the rule are

quoted, followed by explanation, if necessary. A copy of the required LER form (NRC Form

366), LER Text Continuation form (NRC Form 366A), and LER Failure Continuation form (NRC

Form 366B), are shown at the end of this section. The use of LER information and the review

programs associated with LERs are explained in Appendix C.

5.1 LER Reporting Guidelines

This section addresses administrative requirements and provides guidelines for submittal.

Topics addressed include submission of reports, forwarding letters, cancellation of LERs, report

legibility, reporting exemptions, reports other than LERs that use LER forms, supplemental

information, revised reports, and general instructions for completing LER forms.

5.1.1 Submission of LERs

§50.73(d)

"Licensee Event Reports must be prepared on Form NRC 366 and submitted within 30 days

of discovery of a reportable event or situation to the U.S. Nuclear Regulatory Commission,

as specified in §50.4."

An LER is to be submitted (mailed) within 30 days of the discovery date. If a 30-day period

ends on a Satu ray; Sunday, or holiday, reports submitted on the first working day following the

end of the 30 days are acceptable. If a licensee knows that a report will be late or needs an

additional day or so to complete the report, the situation should be discussed with the

appropriate NRC regional office. See Section 2.11 for further discussion of discovery date.

5.1.2 LER Forwarding Letter and Cancellations

The cover letter forwarding an LER to the NRC should be signed by a responsible official.

There is no prescribed format for the letter. The date the letter is issued and the report date

should be the same. Licensees are encouraged to include the NRC resident inspector and the

Institute of Nuclear Power Operations (INPO) in their distribution. Multiple LERs can be

forwarded by one forwarding letter.

99 NUREG-1022, Rev. 1
Cancellationiso LERS submitted~ should be made by letter. The bases for the.---c c .ancellation

should be explained so that the staff can understand and review the reasons supporting the

determination. -The notice of cancellation will be filed and stored with the LER and

acknowledgement made in-various automated data systems.

5.1.3 Report Legibility

§50.73(e)

"The reports and copies that licensees are required to submit to the Commission under the

provisions of this section must be of sufficient quality to permit legible reproduction and

micrographic processing."

No further explanation is necessary.

5.1.4 Exemptions

§50.73(f)

"Upon written request from a licensee including adequate justification or at the initiation of

the NRC staff, the NRC Executive Director for Operations may, by a letter to the licensee,

grant exemptions to the reporting requirements under this section."

Exemptions may be plant specific or generic. However, one of the goals of the LER rule is a

consistent set of reporting requirements that apply to all plants. To minimize inconsistencies in

the reporting, plant-specific exemptions will not be issued unless justified by unique plant

conditions.

5.1.5 Voluntary LERs

Indicate information-type LERS (i.e., voluntary LERs) by checking the "Other" block in Item 11

of the LER form and type 'Voluntary Report" in the space immediately below the block. Also

give a sequential LER number to the voluntary report as noted in Section 5.2.4(5). Because not

all requirements of §50.73(b), "Contents," may pertain to some voluntary reports, licensees

should develop the content of such reports to best present the information associated with the

situation being reported.

See Section 2.9 for additional discussion of voluntary LERs.

NUREG-1022, Rev. 1 100
5.1.6 Supplemental Information and Revised LERs

§50.73(c)

"The Commission may require the licensee to submit specific additional information beyond

that required by paragraph (b) of this section if the Commission finds that supplemental

material is necessary for complete understanding of any unusually complex or significant

event. These requests for supplemental information will be made in writing and the licensee

shall submit, as specified in §50.4, the requested information as a supplement to the initial

LER."

This provision authorizes the NRC staff to require the licensee to submit specific supplemental

information.

If an LER is incomplete at the time of original submittal or if it contains significant incorrect

information of a technical nature, the licensee should use a revised report to provide the

additional information or to correct technical errors discovered in the LER. Identify the revision

to the original LER in the LER number as described in Section 5.2.4(5).

The revision should be complete and should not contain only supplementary or revised

information to the previous LER because the revised LER will replace the previous report in the

computer file. In addition, indicate in the textbhn th LER form the•re6ved r sudippmIentar*

jnfrmtiol~ pacinga a~~tcatfn~ewI helnargInA

If an LER mentions that an engineering study was being conducted, report the results of the

study in a revised LER only if it would significantly change the reader's perception of the course,

significance, implications, or consequences of the event or if it results in substantial changes in

the corrective action planned by the licensee.

Use revisions only to provide additional or corrected information about a reported event. Do not

use a revision to report subsequent failures of the same or like component, except as permitted

in 10 CFR 50.73. Some licensees have incorrectly used revisions to report new events that

were discovered months after the original event because they were loosely related to the

original event. These revisions had different event dates and discussed new, although similar,

events. Report events of this type as new LERs and not as revisions to previous LERs.

5.1.7 Special Reports

There are a number of requirements in various sections of the technical specifications that

require reporting of operating experience that is not covered by 10 CFR 50.73. If LER forms

101 NUREG-1022, Rev. I
are used to submit special reports, check the "Other" block in item 11 of the form and type

"Special Report" in the space immediately below the block. The provisions of §50.73(b) may

not be applicable or appropriate in a special report. Develop the content of the report to best

present the information associated with the situation being reported. In addition, if the LER

form is used to submit a special report, use a report number from the sequence used for LERs.

If an event is reportable both under 10 CFR 50.73 and as a special report, check the block in

Item 11 for the applicable section of 50.73 as well as the "Other" block for a special report. The

content of the report should depend on the reportable situation.

5.1.8 Appendix J Reports (Containment Leak Rate Test Reports)

A licensee must perform containment integrated and local leak rate testing and report the

results as required by Appendix J to 10 CFR Part 50. When the leak rate test identifies a 10

CFR 50.73 reportable situation (see Section 3.2.4 or 3.3.1 of this report), submit an LER and

include the results in an Appendix J report by reference, if desired. The LER should address

only the reportable situation, not the entire leak rate test.

5.1.9 10 CFR Part 21 Reports

10 CFR Part 21, "Reporting of Defects and Noncompliance," as amended during 1991,

encourages licensees of operating nuclear power plants to reduce duplicate evaluation and

reporting effort by evaluating deviations in basic components under the 10 CFR 50.72, 50.73,

and 73.71 reporting criteria. As indicated in 10 CFR 21.2(c) "For persons licensed to operate a

nuclear power plant under Part 50 of this chapter, evaluation of potential defects and

appropriate reporting of defects under §§ 50.72, 50.73. or § 73.71 of this chapter satisfies each

person's evaluation, notification, and reporting obligation to report defects under this part ...." As

discussed in the Statement of Considerations for 10 CFR 21 P4, the only case where a defect in

a basic component of an operating reactor might be reportable under Part 21, but not under §§

50.72, 50.73, or 73.71 would involve Part(s) on the shelf. This type of defect, if it does not

represent a condition reportable under §§ 50.72 or 50.73, might still represent a condition

reportable under 10 CFR Part 21.

For an LER, if the defect meets one of the criteria of 10 CFR 50.73, check the applicable

paragraph in Item 11 of NRC Form 366 (LER Form). Licensees are also encouraged to check

the "Other" block and indicate "Part 21" in the space immediately below if the defect in a basic

component could create a substantial safety hazard. The wording in Item 16 ("Abstract") and

Item 17 ("Text") should-state that the report on stitutes_ a Part 211 notificattion. If the defect Ns

applicable to-other facilities at a multi-Unit site,- a single LER may be used- by. indilcating the

other involved facilities in Item 8 on the LER For1i

04) 56 FR 36081, July 31, 1991.

NUREG-1022, Rev. I 102
5.1.10 Section 73.71 Reports

Submit events or conditions that are reportable under 10 CFR 73.71 using the LER forms with

the appropriate blocks in Item 11 checked. If the report contains safeguards information as

defined in 10 CFR 73.21, the LER forms may still be used, but should be appropriately marked

in accordance with 10 CFR 73.21. Include safeguards and security information only in the

narrative and not in the abstract. In addition, the text should clearly indicate the information that

is safeguards or security information. Finally, the requirements of §73.21(g) must be met when

transmitting safeguards information. F .itional-guidelines on 10 CFR 73.1 reporting tee

Regulatory Guide 5';62, Revision 1, "Reportin~g of Safeguards Events," November -1987;

NUREG-1 304, "Reporting of Safeguards Events," February1 1988; and Generic Letter 91 -03,

MReporting of Safe~guards Events," March 6, 1991.

If the LER contains proprietary information, mark K appropiiately in item 17 (text) on of the LER

form. Include proprietary information only in the narrative and not in the abstra!t. In addition,

indicate clearly in the narrative the information that is proprietary. Finally, the requirements of

.g.90(0) must Vie-Me-t- whe transmittin prpita ryionaton.

5.1.11 Availability of LER Forms

The NRC will provide LER forms (i.e., NRC Forms 366, 366A, and 366B) free of charge.

Copies may be obtained by writing to the NRC Information and Records Management Branch,

Office of the Chief Information Officer, US Nuclear Regulatory Commission, Washington, DC

20555. Electronic versions are also available. Licensees are encouraged to use these forms to

assist the NRC's processing of the reports.

5.2 LER Content Requirements and Preparation Guidance

Licensees are required to prepare an LER for those events or conditions that meet one or more

of the criteria contained in §50.73(a). Paragraph 50.73(b), "Contents," specifies the information

that an LER should contain with further explanation when appropriate.

In 1986, the NRC decided to use an optical character reader (OCR) to read LER abstracts into

NRC LER data bases (IE Information Notice No. 86-08, "Licensee Event Report (LER) Format

Modification," February 3, 1986). At that time, licensees were asked to help reduce the number

of errors incurred by the OCR as a result of incompatible print styles by using OCR-compatible

typography for preparing LERs. Therefore, certain limitations have been placed on the use of

type styles and symbols for the abstract and text of the LERs. These limitations are listed

below. (See the Information Notice for details.)

Type Styles:

* Prestige Elite (12 pitch)

0 Letter Gothic (12 pitch)

0 OCR-B (12 pitch)

0 Courier 12 (12 pitch)

0 Elite (12 pitch)

103 NUREG-1022, Rev. 1
0 Courier 10 (10 pitch)

0 OCR-A (10 pitch)

* Prestige Pica (10 pitch)

0 Prestige Pica (10 pitch)

In addition, the following proportional space type-styles can be read: Madeleine, Cubic, Bold,

and Title.

It is suggested that output be on typewriter or formed character (letter-quality or near letterquality) printer (e.g., daisy wheel, laser, ink-jet).

It is suggested that output have an uneven right margin (i.e., we suggest that you n2t right

justify output).

It is suggested that text of the abstract be kept at least Y2-inch inside the border on all sides of

the area designated for the abstract on the LER form. Text runnin4'gintoth• b -rder can interm

with scanning the d6cmen6t1

It is suggested that you do nog use underscore, do not use bold print, do not use Italic print

style, do not end any lines with a hyphen and do not use paragraph indents. Instead, print copy

single space with a blank line between paragraphs.

Limitations on the use of symbols in the textual areas:

" Spell out the word "degree."

* Use </= for "less than or equal to."

* Use >/= for "greater than or equal to."

" Use +/- for "plus or minus."

* Spell out all Greek letters.

Do not use exponents. A number should either be expressed as a decimal, spelled out, or

preferably designated in terms of "E" (E field format). For example, 4.2 x 10- could be

expressed as 4.2E-6, 0.0000042, or 4.2 x 10(-6).

Define all, abbreviations and acronyms In both the tex and: the abstract ahd expfa1r4.

component designators the first time they are used (e~g., the emergency service water pump 1.

SW-P-1A).,

NUREG-1022, Rev. 1 104
5.2.1 Narrative Description or Text (NRC Form 366A, Item 17)

(1) General

§50.73(b)(2)(i)

The LER shall contain: "A clear, specific, narrative description of what occurred so that

knowledgeable readers conversant with the design of commercial nuclear power plants,

but not familiar with the details of a particular plant, can understand the complete event."

There is no prescribed format for the LER text; write the narrative in a format that most clearly

describes the event. AtuhS.`),finres tie I#nTiati6n t hould tec l61tud6-d, 11 is

,t i ,.tonded;94, ei,,ntline of the te4 fco•mt. ,'After the narrative is written, however, review the

appropriate sections of §50.73(b) to make sure that applicable subjects have been adequately

addressed. Wis helpful to use headings to Improve-radabIlity. For examplesome LERs

employ major headings-such as event description,, safety consequences, corrective, actions,

prnd previous similar events- and subheadings such 'as initialrc .on -ditions, dates and times,.event

classification, systenms status,- event or condition causes,-falure modes, method -f dis~covery,

component information, immediate corrective actions, and actions to prevent recurrence.

Explain exactly what happened during the entire event or condition, including how systems,

components, and operating personnel performed. Do not cover specific hardware problems in

excessive detail. Describe unique characteristics of a plant as well as other characteristics that

influenced the event (favorably or unfavorably). Avoid using plant-unique terms and

abbreviations, or, as a minimum, clearly define them. The audience for LERs is large and does

not necessarily know the details of each plant.

Include the root causes, the plant status before the event, and the sequence of occurrences.

Describe the event from the perspective of the operator (i.e., what the operator saw, did,

perceived, understood, or misunderstood). Specific information that should be included, as

appropriate, is described in paragraphs 50.73(b)(2)(ii), (b)(3), (b)(4), and (b)(5) of the rule and

separately in the following sections.

If several cngi,,,rod of3 systems actuate during an event, describe all

aspects of the complete event, including all actuations sequentially, and those aspects that by

themselves would not be reportable. For example, if a f iig component failure

(generally not reportable) occurs following a reactor scram (reportable), describe the

component failure in the narrative of the LER for the reactor scram. The" b; 

pA enn fin~~ nptt~~•'nmpa.tIt is necessary to discuss the

performance and status of $fi equipment important for defining and understanding what

happened and for determining the potential implications of the event.

Paraphrase pertinent sections of the latest submitted safety analysis report (SAR) rather than

referencing them because not all organizations or individuals have access to SARs. Extensive

cross-referencing would be excessively time consuming considering the large number of LERs

105 NUREG-1022, Rev. 1
and large number of reviewers that read each LER. Ensure that each applicable component's

safety-significant effect on the event or condition is clearly and completely described.

Do not use statemensuch as "this event is not significant with respect to the health and safety

of the public" without explaining the basis for the conclusion;

§50.73(b)(2)(ii)(A)

The narrative description must include: "Plant operating conditions before the event."

Describe the plant operating conditions such as power level or, if not at power, describe mode,

temperature, and pressure that existed before the event.

§50.73(b)(2)(ii)(B)

The narrative description must include: "Status of structures, components, or systems that

were inoperable at the start of the event and that contributed to the event."

If there were no structures, systems, or components that were inoperable at the start of the

event and contributed to the event, so state. Otherwise, idetifyý SSCs that were inoperable

§50.73(b)(2)(ii)(C)

The narrative description must include: "Dates and approximate times of occurrences."

For a transient or ESF actuation event, the event date and time are the date and time the event actually occurred. If the event is a discovered condition for which the occurrence date is not

known, the event date should be specified as the discovery date. However, a discussion of the

best estimate of the event date and its basis should be provided in the narrative. For example,

if a design deficiency was identified on March 27, 1997 that involved a component installed during refueling in the spring of 1986, and only the discovery date is known with certainty, the

event date should be specified as the discovery date. A discussion should be provided that

describes, based on the best information available, the most likely time that the design flaw was

introduced into the component (e.g, by manufacturer or by plant engineering prior to

procurement). The length-of time that the cOmponent Was in service should also be =rovided

(L~e., When it was installed~~i

NUREG-1022, Rev. 1 106
Discuss both the discovery date and the event date if they differ. If an LER is not submitted

within 30 days from the event date, explain the relationship between the event date, discovery

date, and report date'in the narrative. See Section 2.11 for further discussion of discovery date.

Give dates and approximate times for all major occurrences discussed in the LER (e.g.,

discoveries; immediate corrective actions; systems, components, or trains declared inoperable

or operable; reactor trip; actuation and termination of equipment operation; and stable

conditions achieved). inparticuar, fo stndby pumps •• and m1erency generators, indicate the

length"o t~imfeof- 01-9raion-and any -intermittent periods Of Shutdown or inoperability during the

event. Include an estimate of the time and date of failure of systems, components, or trains if

different from the time and date of discovery. A chronologYmay be used to clarify teltimigof

personnel, and equipm#tent c-5ions.

For equipment that was inoperable at'e start ofthe eV Provide an estimate of the time the

equipment became inoperable and the last time the equipmenit was known to be operable.

Indicate the-basis for this conclusion (e.g., a test was successfully run or the equipment was

operating). ~Forequipmnent that tailed, provide the fa-ilure-time~and the last time the equipment

was known to be operable.Also provide the basis for the lastqtime known operable.

Components such as valves and snubbers may be tested over a period of several weeks.

During this period, a number of inoperable similar components may be discoveredP' In such

cases, similar failures that are reportable and that are discovered during a single test program

within the 30 days of discovery of the first failure may be reported as one LER. For similar

failures that are reportable under Section 50.73 criteria and that are discovered during a single

test program or activity, report all failures that occurred within the first 30 days of discovery of

the first failure on one LER. However, the 30-day clock starts when the first reportable event is

discovered. State in the LER text (and code the information in Items 14 and 15) that a

supplement to the LER will be submitted when the test is completed. Submit a revision to the

original LER when the test is completed. Include all the failures, including those reported in the

original LER, in the revised LER (i.e., the revised LER should stand alone).

(3) Failures and Errors

§50.73(b)(2)(Ii)(D)

The narrative description must include: "The cause of each component or system failure or

personnel error, if known."

Include the roo causeiWyldentlfied for each -coMplonent or syst.em failure(ofat)rpesne

pfror. Contiributing facto-r-s may be discussed as appropriate: For example, @.valve stem

r Note that inoperable similar components might indicate common cause failures of

independent trains or channels, which are reportable under §50.73(a)(2)(vii); see Section 3.3.4

for further discussion.

107 NUREG-1022, Rev. 1
breaking couild ha~e Ween caused by a limit; switch that had been impr~oderly adjuted-durdn

maintenance; in-thi case, the roof cause might be determined to be personnel-error and

additional discus -sion could focus on -the limit switch adjustment. If the personnel error ]s

determined to have -been caused by deficient procedures or inadequate personnel training, thi-s

should be explainedi.

If the cause of a failure canntdtbe readilydterOmined and tieivsiaintcniuntha

LER should indicate what additional investigation is planned. -A supplemental LER should be

submitted foillowing- the7 additional Investigation -if substantial information Is I Ide -ntified that-wo~id

significantly change a readers perception of the course -or consequences of the event- or if

there are substantial change's in~ the- -corrective actions planned by the 1icenseei

§50.73(b)(2)(ii)(E)

The narrative description must include: "The failure mode, mechanism, and effect of each

Include the-fairem~ode, mechanism- (immediate cause)-, and -effect-Of eaci* failedcompOftentinthe narrative. The~ effect of the failure on safety_ systems and functions should be fully. described. Identify the specific piece part that failed and the specific trains and -systerYMi

rendered inoperable or degraded.- Iden~tify all dependent systems rendered inop~rabeor

degraded. Indicate whether redfundant trains were-operable7 and avatlablwi

If the equipment is degraded, but not failed,- describe the degradation and itseffects and

indicate why the equiipment would still perform its intended functionY

§50.73(b)(2)(ii)(F)

The narrative description must include: "The Energy Industry Identification System

component function identifier and system name of each component or system referred to in

the LER.

(1) The Energy Industry Identification System is defined in: IEEE Std 803-1983 (May 16,

1983) Recommended Practice for Unique Identification in Power Plants and Related

Facilities-Principles and Definitions.

(2) IEEE Std 803-1983 has been approved for incorporation by reference by the Director

of the Federal Register.

A notice of any changes made to the material incorporated by reference will be published in

the Federal Register. Copies may be obtained from the Institute of Electrical and

Electronics Engineers, 345 East 47th Street, New York, NY 10017. IEEE Std 803-1983 is

available for inspection at the NRC's Technical Library, which is located in the Phillips

(Continued on next page)

NUREG-1022, Rev. 1 108
§50.73(b)(2)(ii)(F) (Continued)

Building, 7920 Norfolk Avenue, Bethesda, Maryland; and at the Office of the Federal

Register, 1100 L Street, NW, Washington, DC."

Note: The NRC library Is Anow tocaed. In the Two White Flin INot u1dinh114Rcive

Pike, Rockville, Maryland.

The system name may be either the full name (e.g., reactor coolant system) or the two-letter

system code (such as AB for the reactor coolant system). However, when the name is long

(e.g., low-pressure coolant injection system), the system code (e.g., BO) should be used. If the

full names are used, The Energy Industry Identification System (EIIS) component function

identifier and/or system identifier (i.e., the two letter code) should be included in parentheses

following the first reference to a component or system in the narrative. The component function

identifiers and system identifiers need not be repeated with each subsequent reference to the

same component or system.

q yete ie-Iif an ' n~eij tt e thi 411y-eta-OM repi ile 4j9P46 

p, compjonenrwth~in thIe -scope-of th~e Equipment- erformanrce7and InformatiornExchange-(EPIX)

~ytem i nv v the~te system-and train desgnation should be coflistenyt with-the: IIS used in

EPIX.

§50.73(b)(2)(ii)(G)

The narrative description must include the following specific information as appropriate for

the particular event: "For failures of components with multiple functions, include a list of

systems or secondary functions that were also affected."

No further explanation is necessary.

§50.73(b)(2)(ii)(H)

The narrative description must include: "For failure that rendered a train of a safety system

inoperable, an estimate of the elapsed time from the discovery of the failure until the train

was returned to service."

N165 further e~xplana~tiohnisriecessary.

109 NUREG-1022, Rev. 1
§50.73(b)(2)(ii)(1)

The narrative description must include: 'The method of discovery of each component or

system failure or procedural error."

Explain how each component failure, sy*stem failure-, p~ersonnel e*rro*r*, or procedural deficiencY

was discovered. Examples include revi~ewing surveillance procedures or results -of s Urveiflaned

tests, pre-startup valve lineup check, performring quarterly maintenance, plant walkdown, etc.

§50.73(b)(2)(ii)(J)

The narrative description must include the following specific information as appropriate for

the particular event:

"(1) Operator actions that affected the course of the event, including operator errors,

procedural deficiencies, or both, that contributed to the event.

(2) For each personnel error, the licensee shall discuss:

(i) Whether the error was a cognitive error (e.g., failure to recognize the actual plant

condition, failure to realize which systems should be functioning, failure to recognize the

true nature of the event) or a procedural error;

(ii) Whether the error was contrary to an approved procedure, was a direct result of an

error in an approved procedure, or was associated with an activity or task that was not

covered by an approved procedure;

(iii) Any unusual characteristics of the work location (e.g., heat, noise) that directly

contributed to the error, and

(iv) The type of personnel involved (i.e., contractor personnel, utility-licensed operator,

utility non-licensed operator, other utility personnel)."

Human performance often influences the outcome of nuclear power plant events. Human error

is known to contribute to more than half of the LERs. The LER rule identifies the types of

reactor events and problems that are believed to be significant and useful to the NRC in its

effort to identify and resolve threats to public safety. It is designed to provide the information

necessary for engineering studies of operational anomalies and trends and patterns analysis of

operational occurrences includlg human performance*.

Generally, the criteria of Section 50.73(b)(2)(i) require a clear, specific narrative so that

knowledgeable readers can understand the complete event. Further, the criteria of Section

50.73(b)(2)(ii)(J) require a description of (1) operator actions that affected the course of the

event and (2) for each personnel error, additional specific information as detailed in the rule. 1rW

order to tuppmran Understanfding of human pe-Kbrifm~anh-de--i ss Ue6sieJl ate d * toMe -e--ve n! id

the factors- dfiscussed below to the- extent; they apply._ For example,i Jan operator ertorthat

affected the course-of the event wasdueto a proce.dural -problem, indicate th-e na-ture of thd

procedural problem such as m Issing procedure,- -pro cedure inadequate -due to techn-icmi

deficiency;ý etc.,

NUREG-1022, Rev. I 110
Personnel errors aind human -performance related issues-rlay bein the areas of procedures,

training, comnmunication, human enginee~ring, management, and supervision. For example, in

the area of proceduires, -errors might-be due to missing proceduires,, procedures which are

Inadequate due to technical or human factors -deficiencies,- or which have not been maintained

turrent. -In the area of training, errors m~ay be the result of a failure to provide training, having

Provided inadequate training, or as the result of training (such- assimulator training or'on-the-job

Jraining) that does-not provide an'environment comparable to-that in the-plant.

Pommunications errors may be due to inadequate, untimety.,rnisunderstood, or mis~sing2q

pommunlcation or due to the quality of the communication equipment. Human engineering

Issues include those-related t -o the interface or lack thereof between -the human and the

trachine (such as-size, shape,- location, function-or content of displays, controls, equipment or

labels)-as well as environmental is -sues such as lighting, temperature, noise, radiation and work

prea layout. Management errors Might be due to management expectations, corrective -actions,

toot -cause-determinations, or audits which- are iniadequate~ unttimely or missing. In the area of

stupervision, errors may-be the result of a -lack of supervision, inadequate supervision, job

Otaffing, overti~me- -scheduling ~and plannin., work practices (such as briefings, logs, work

ýackage s-- -team work, decision making, and hiousekeeping)-or because of inadequate

yeiification. awarene ss or seif-checking

§50.73(b)(2)(ii)(K)

The narrative description must include: "Automatically and manually initiated safety system

Tihe LER should' include a disc~s~ion of eac specific sytterfi that actuated or failfed to actulate.'.

Oo not flmi -t the discussion to ESFs. Indicate the.specific equipment that-actuated or should

have actuated, -by train, com patible with.EP IX train definitions (e.g., auxiliary feedwater train' B).

indicate Whether or not the equipment bperated successdflly;. [ §50.73(b)(2)(ii)(L)

The narrative description must include: 'The manufacturer and model number (or other

identification) of each component that failed during the event."

TUhe maniufatturer and model number (or other identification, uch as type, size, or manufacture

tlatey .also should b-gvnfor"each component found-failed during the course of the event. An

Oxample of other identification could be (for a pipe rupture) size, schedule, or material

tom Ipositionill NUREG-1022, Rev. 1
(3) Assessment of Safety Consequences

§50.73(b)(3)

The LER shall contain: "An assessment of the safety consequences and implications of the

event. This assessment must include the availability of other systems or components that

could have performed the same function as the components and systems that failed during

the event."

Give a summary assessment of the actual and potential safety consequences and implications

of the event, including the basis for submitting the report. Evaluate the event to the extent

necessary to fully assess the safety consequences and safety margins associated with the

event.

Include an assessment of the event under alternative conditions if the incident would have been

more severe (e.g., the plant would have been in a condition not analyzed in its latest SAR)

under reasonable and credible alternative conditions, such as a different operating mode. For

example, if an event occurred while the plant was at low power and the same event could have

occurred at full power, which would have resulted in considerably more serious consequences,

this alternative condition should be assessed and the consequences reported.

Reasonable and credible alternative conditions may include normal plant operating conditions,

potential accident conditions, or additional component failures, depending on the event. Normal

alternative operating conditions and off-normal conditions expected to occur during the life of

the plant should be considered. The intent of this section is to obtain the result of the

considerations that are typical in the conduct of routine operations, such as event reviews, not

to require extraordinary studies.

(4) Corrective Actions

§50.73(b)(4)

The LER shall contain: "A description of any corrective actions planned as a result of the

event, including those to reduce the probability of similar events occurring in the future."

Incuiud6wheher 6dieacina-fT Of fie~ 2bempj MWM&DIs rep&atrl

relcmn atosa ela actions that wiJI reduce theprobability-Ofa sirnitarevenj .occurring 1in-theT future. For excample, "the pump was 1re .p aired and a discusionr of the event

the intrument-before the step in-Wihte vn a ntaed.1

NUREG-1022, Rev. I 112
In addition to a description of any corrective actions planned as a result of the event, describe

corrective actions on similar or related components that were done, or are planned, as a direct

result of the event. For example, if pump 1 failed during an event and required corrective

maintenance and that same maintenance also was done on pump 2, so state.

If a study was conducted, and results are not available within the 30-day period, report the

results of the study in a revised LER if they result in substantial changes in the corrective action

planned. (See Section 5.1.6 for further discussion of submitting revised LERs.)

(5) Previous Occurrences

§50.73(b)(5)

The LER shall contain: "Reference to any previous similar events at the same plant that are

known to the licensee."

The term "previous occurrences" should include previous events or conditions that involved the

same underlying concern or reason as this event, such as the same root cause, failure, or

sequence of events. For infrequent events such as fires, a rather broad interpretation should

be used (e.g., all fires and, certainly, all fires in the same building should be considered

previous occurrences). For more frequent events such as ESF actuations, a narrower definition

may be used (e.g., only those scrams with the same root cause). The intent of the rule is to

identify generic or recurring problems.

The licetnee should ue ent 6n0e1r6 JUdgtment to de Win ehiow far brak in time sb to 6 tc oe ent, a

reasc nably cornpete picture aiofne curri nt problt m Theintent is -to bable to see a pattern in

recurring eventsmoratheroha c shplete 1L0 20year history -of the system. Ifthe

e~vent was a high-frequency~ type of event, -2 years back 'mayjbe -more than sufficient.

Include the LER nuhber(s),e if any of previious similar oeents. Previous shmeiar evethate m ny t

necessarily limited geto events, reported aphs, In an ot prevos isinAreventh ae nnarrv so

state. If any earlier events, in retrospect, were significant in relation to the subject event,

discuss why prior corrective action did not prevent recurrence.

(6) LER Text Continuation Sheet (NRC Form 3666)

Use one or more additional text continuation sheets of the LER Form 366A to continue the

narrative, if necessary. There is no limit on the number of continuation sheets that may be

included.

Drawings, figures, tables, photographs, and other aids may be included with the narrative to

help readers understand the event. If possible, provide the aids on the LER form (i.e., NRC

Form 366A). In addition, care should be taken to ensure that drawings and photographs are of

sufficient quality to permit legible reproduction and micrographic processing. Avoid oversized

drawings (i.e., larger than 8 % x 11).

113 NUREG-1022, Rev. I
5.2.2 Abstract (NRC Form 366, Item 16)

§50.73(b)(1)

The LER shall contain: "A brief abstract describing the major occurrences during the event,

including all component or system failures that contributed to the event and significant

corrective action taken or planned to prevent recurrence."

Provide a brief abstract describing the major occurrences during the event, including all actual

component or system failures that contributed to the event, all relevant operator errors or

violations of procedures, the root cause(s) of the major occurrence(s), and the corrective action

taken or planned for each root cause. If space does not permt describing failures at least

indicaWt Whether or nat failures occurred Limit the abstract to 1400 characters (including

spaces), which is approximately 15 lines of single-spaced typewritten text. Do not use EIIS

component function identifiers or the two-letter codes for system names in the abstract.

It ~ o ~~ptbk do~rl~ h~ ntit Ccntin hostfbt spade-._ Ho~cvet, the deodoibmi.n 3

the event sheuld be sufflieintly detailed so that a k~nowledgeable reader ean understand thol

qempleta eve~nt. Few fepeftable cvents will be se sim pliiatie thaet they gan be Eldequoel desritocl in 1400 eharera0_ -'

The abstract is generally included in the LER data base to give users a brief description of the

event to identify events of interest. Therefore, if space permits, provide the numbers of other

LERs that reference similar events in the abstract.

As noted in Section 5.1.10, do not include safeguards, security, or proprietary information in the

abstract.

5.2.3 Other Fields on the LER Form

(1) Facility Name (NRC Form 366. Item 1)

Enter the name of the facility (e.g., Indian Point, Unit 1) at which the event occurred. If the

event involved more than one unit at a station, enter the name of the nuclear facility with the

lowest-nuclear unit number (e.g., Three Mile Island, Unit 1).

(2) Docket Number (NRC Form 366. Item 2)

Enter the docket number (in 8-digit format) assigned to the unit. For example, the docket

number for Yankee-Rowe is 05000029. Note the use of zeros in this example.

(3) Page Number (NRC Form 366. Item 3)

Enter the total number of pages included (including figures and tables that are attached to Item

17 Text) in the LER package. For continuation sheets, number the pages consecutively

NUREG-1022, Rev. I 114
beginning with page 2. The LER form, including the abstract and other data is pre-numbered

on the form as page 1 of.._,

(4) Title (NRC Form 366. Item 4)

The title should include a concise description of the principal problem or Itsue tociated With

!he even, the rM~ c~ut te result (Why-theeVet wa -required to be -reportedy,_ and th~e link

between them, if possibte.- It isoften easier toform the title -after writing the assessment of the

Ovent because the in~formation lslear~y at hand.

"Licensee Event Report" should not be used as a title. The title "Reactor Trip"is considered

inadeqUate6, because the root ca-0-ste-a endthe71inK-bVe-hterhe root cause and the result are

Mnissing. The title "'Personnel Error Causes Reactor Trip"' is considered inadequate because of

thei innui merabl e ways in which ' a pesoncould cause-a m ctbrrij 7'"Technician tnadvertently'

injected Signal Reultingirn a-~Reactor Tdr wouldbee 0 betertitle.

(5) Event Date (NRC Form 366. Item 5)

Enter the date on which the event occurred in the pight spaces provided. There are two spaces

for the month, two for the day, and four for the year, in that order. Use leading zeros in the first

and third spaces when appropriate. For example, June 1, 1987, would be properly entered as

06011987.

If the date on which the event occurred cannot be clearly defined, use the discovery date. See

Section 2.11 of this report for further discussion of discovery date.

(6) Report Number (NRC Form 366. Item 6)

The LER number consists of three parts: (a) the Iou digits of the event year (based on event

date), (b) the sequential report number, and (c) a revision number. The numbering system is

shown in the diagram below; the event occurred in the year 1991, it was the 45th event of that

year, and the submittal was the 1st revision to the original LER for that event.

Event Sequential Revision

Year Report Number Number

J191 - 045 - 01

Event Year: Enter the r digits. The event year should be based on the event date (Item 4).

Sequential Report Number: As each reportable event is reported for a unit during the year, it is

assigned a sequential number. For example, for the 15th and 33rd events to be reported in a

given year at a given unit, enter 015 and 033, respectively, in the spaces provided. Follow the

guidelines below to ensure consistency in the sequential numbering of reports.

0 Each unit should have its own set of sequential report numbers. Units at multi-unit sites

should not share a set of sequential report numbers.

115 NUREG-1022, Rev. 1
* The sequential number should begin with 001 for the first event that occurred in each

calendar year, using leading zeros for sequential numbers less than 100.

* For an event common to all units of a multi-unit site, assign the sequential number to the

lowest numbered nuclear unit.

* If a sequential number was assigned to an event, and it was subsequently determined

that the event was not reportable, a "hole" in the series of LER numbers would result.

The NRC would prefer that licensees reuse a sequential number rather than leave holes

in the sequence. A sequential LER number may be reused even if the event date was

later than subsequent reports.

If the licensee chooses not to reuse the number, write a brief letter to the NRC noting that "LER

number xxx for docket O500OXXX will not be used."

Revision Number. The revision number of the original LER submitted is 00. The revision

number for the first revision submitted should be 01. Subsequent revisions should be numbered

sequentially (i.e., 02, 03, 04).

(7) Report Date (NRC Form 366. Item 7)

Enter the date the LER is submitted to the NRC in the eht spaces provided, as described in

Section 5.2.4(4) above.

(8) Other Facilities (NRC Form 366. Item -)

When a situation is discovered at one unit of a facility that applies to more than the one unit,

submit a single LER. LER form items 1, 2, 6, 9, and 10 should refer to the unit primarily

affected, or, if both units were affected approximately equally, to the lowest numbered nuclear

unit.

The intent of the requirement is to name the facility in which the primary event occurred,

whether or not that facility is the lowest numbered of the facilities involved. The automatic use

of the lowest number should only apply to cases where both units are affected approximately

equally. Item 8 only should indicate the other unit(s) affected. The abstract and the text should

describe how the event affected all units.

Enter the facility name and unit number and docket number (see Sections 5.2.4(1) and 5.2.4(2)

for format) of any other units at that site that were directl affected by the event (e.g., the event

included shared components, the LER described a tornado that threatened both units of a twounit plant).

(9) Operating Mode (NRC Form 366. Item 9)

Enter the operating mode of the unit at the time of the event as defined in the plant's technical

specifications in the single space provided. For plants that have operating modes such as hot

NUREG-1022, Rev. 1 116
shutdown, cold shutdown, and operating, but do not have numerical operating modes (e.g.,

Mode 5), place the letter N in Item 9 and describe the operating mode in the text.

(10) Power Level (NRC Form 366. Item 10)

Enter the percent of licensed thermal power at which the reactor was operating when the event

occurred. For shutdown conditions, enter 000. For all other operating conditions, enter the

correct numerical value (estimate power level if it is not known precisely), using leading zeros

as appropriate (e.g., 009 for 9-percent power). Significant deviations in the operating power in

the balance of plant should be clarified in the text.

(11) Reporting Requirements (NRC Form 366. Item 11)

Check one or more blocks according to the reporting requirements that apply to the event. A

single event can meet more than one reporting criterion. For example: if as a result of

sabotage, reportable under §73.71(b), a safety system failed to function, reportable under

§50.73(a)(2)(v), and the net result was a release of radioactive material in a restricted area that

exceeded the applicable license limit, reportable under §20.2203(a)(3)(i), prepare a single LER

and check the three boxes for paragraphs 73.71(b), 50.73(a)(2)(v), and 20.2203(a)(3)(i).

In addition, an event can be reportable as an LER even if It does not meet any of the criteria of

10 CFR 50.73. For example, a case of attempted sabotage (§73.71(b)) that does ngt result in

any consequences that meet the criteria in 50.73 can be reported using the "Other" block. Use

the "Other" block if a reporting requirement other than those specified in item 11 was met.

Specifically describe this other reporting requirement in the space provided below the "Other"

block and in the abstract and text.

(12) Licensee Contact (NRC Form 366. Item 12)

§50.73(b)(6)

The LER shall contain: "The name and telephone number of a person within the licensees

organization who is knowledgeable about the event and can provide additional information

concerning the event and the plant's characteristics."

Enter the name, position title, and work telephone number (including area code) of a person

who can provide additional information and clarification for the event described in the LER.

(13) Component Failures (NRC Form 366. Item 13)

Enter the appropriate data for each component failure described in the event.

A failure is defined as the termination of the ability of a component to perform its required

function. Unannounced failures are not detected until the next test; announced failures are

detected by any number of methods at the instant of occurrence.

117 NUREG-1022, Rev. 1
If multiple components of the same type failed and all of the information required in Item 13

(i.e., cause, system, component, etc.) was the same for each component, then only a single

entry is required in Item 13. Clearly define the number of components that failed in the abstract

and text.

The component information elements of this item are discussed below.

Cause: Enter the cause code as shown below. If more than one cause code is applicable,

enter the cause code that most closely describes the root cause of the failure.

Cause

Code Classification and Definition

A P is assigned to failures attributed to human errors. Classify

errors made because written procedures were not followed or because

personnel did not perform in accordance with accepted or approved practice as

personnel errors. Do not include errors made as a result of following incorrect

written procedures in this classification.

B Design. Manufacturing. Construction/Installation is assigned to failures

reasonably attributed to design, manufacture, construction, or installation of a

system, component, or structure. For example, include failures that were traced

to defective materials or components otherwise unable to meet the specified

functional requirements or performance specifications in this classification.

C External Cause is assigned to failures attributed to natural phenomena. A typical

example would be a failure resulting from a lightning strike, tornado, or flood.

Also assign this classification to man-made external causes that originate off site

(e.g., an industrial accident at a nearby industrial facility).

D Defective Procedure is assigned to failures caused by inadequate or incomplete

written procedures or instructions.

E Management/Quality Assurance Deficiency is assigned to failures caused by

inadequate management oversight or management systems (e.g., major

breakdowns in the licensee's administrative controls, preventive maintenance

program, surveillance program, or quality assurance controls, inadequate root

cause determination, inadequate corrective action).

X Qther is assigned to failures for which the proximate cause cannot be identified

or which cannot be assigned to one of the other classifications.

System: Enter the two-letter system code from Institute of Electrical and

Electronics Engineers (IEEE) Standard 805-1984, "IEEE Recommended Practice

for System Identification in Nuclear Power Plants and Related Facilities," March

27, 1984. Copies may be obtained from the Institute of Electrical and Electronics

Engineers, 345 East 47th Street, New York, NY 10017.

NUREG-1022, Rev. I 118
Comoonent: Enter the applicable component code from IEEE Standard

803A-1983, "IEEE Recommended Practice for Unique Identification in Power

Plants and Related Facilities - Component Function Identifiers."

Component Manufacturer: Enter the four character alphanumeric reference

M61 eemputerioer fistin~g of me-nufaet- r-er ec mitained on1 JNPG's

ccrnputer. _Designate menufa_ e.~r tha st -net ineluded in the list ~as-4V)D. if

the manufacturer-s one-used inEPpXuse the manufacturer name as it app .e ars

InEPIX.1

Reportable to EPEI: Enter a "Y" if the failure is reportable to EPIX and an "N" if it

is not reportable.

Include in the LER text and in item 13 of the LER Form any component failure

involved in the event, not just components within the scope of •PIX or EIIS.

Failure Continuation Sheet (NRC Form 366D): If more than four failures need to

be coded, use one or more of the failure continuation sheets (NRC Form 366B).

Code the entries in Items 1, 2, 3, and 6 of the failure continuation sheet to match

entries of these items on the initial page of the LER. Complete item 13 in the

same manner as item 13 on the basic LER form. Do not repeat failures coded

on the basic LER form on the failure continuation sheet. Place any failure

continuation sheets after any text continuation sheets and include those sheets

in the total number of pages for the LER.

(14) Supopemental Report (NRC Form 366. Item 14)

Check the 'Yes" block if the licensee plans to submit a followup report. For example, if a failed

component had been returned to the manufacturer for additional testing and the results of the

test were not yet available when the LER was submitted, a followup report would be submitted.

(15) Expected Submission Date of Supplemental Report (NRC Form 366. Item 15)

Enter the expected date of submission of the supplemental LER, if applicable. See Section

5.2.4(4) for the proper date format. The expected submission date Is tret-ng-date;i•t

Is not a regula-t ory colmlmiftment.:

119 NUREG-1022, Rev. I
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES MMJDD/YYYY

(MM-YYYY) Estimated burden per response to comply with this mandatory Information

collecton request 50 hr&. Reportedlessons learned am incorporated Into the

licnsiig rocss ndfed back to Indiustry. Foiward conmmentsrgring LICENSEE EVENT REPORT (LER) buren esme to the Information and Records a Branc crh O

F33), U.S. Nuclear Regulatory CommissioWashingo 20555-0001,and

(See reverse for required number of to the Paperwork Reduction Project (3150-0104). Office of Management and

Budget. Washington, DC 20503. fa documentusedto imposeaninformation

digits/characters for each block) collection does not display a currently valid OMB control number, tlhe NRC

may not conduct or sponsor, and a person Is not required to respond to. the

Information collection.

FACIUTY NAME (1) DOCKET NUMBER (2) PAGE (3)

05000 1 OF

TITTLE (4)

EVENT DATE (51 LER NUMBER (6I REPORT DATE (7) OTHER FACILITIES INVOLVED (8S

MONTH DAY I YEAR YEAR I SEQUENTIALu I REVISION MONTH DAY YEAR FACILITY NAME DOCKET NUMBER

NUMBER NUMBER DA05000

i FACIUTY NAME DOCKET NUMBER

05000

OPERATING 1 THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 1: (Check one or more) (111

MODE (9)1 20.2201(b) 20.2203(a$12)(v) 50.73(al2)(i) 50.73(a)12)(viii)

POWER 1 20.2203(a)(1) 20.2203(a)13)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x4

LEVEL (10) 20.2203(a)12)(i) 20.22031a)(31(ii) 50.73(a)(2)(iii) 73.71

_______ - ::20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(v) OTHER

20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v) Specify In Abstract below

NAME .120.2203(a)(2)iiv) 50.36(1)12) 50.73(a)(2)(vii) or in NRIC Form 36bA

LICENSEE CONTACT FOR THIS LER 112)

NAVE TELEPHONE NUMBER CIndude Area Code)

COMPLETE ONE LINE FOR FACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (131

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO EPIX TO EPIX

SUPPLEMENTAL REPORT EXPECTED 114) EXPECTED MONTH DAY YEAR

(if yes, complete EXPECTED SUBMISSION DATE). NO DATE (15 1

ABSTRACT (lUmit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

NUREG-1022, Rev. 1 120

NRC FORM 3686 (MM-YYYY)
REQUIRED NUMBER OF DIGITSICHARACTERS

FOR EACH BLOCK

BLOCK NUMBER OF

NUMBER DIGITSICHARACTERS TITLE

1 UP TO 46 FACILITY NAME

28 TOTAL DOCKET NUMBER 2_ 3 IN ADDITION TO 05000

3 VARIES PAGE NUMBER

4 UP TO 76 TITLE

8 TOTAL

5 2 FOR MONTH EVENT DATE

2 FOR DAY

4 FOR YEAR

9 TOTAL

6 3 FOR SEQUENTIAL 4FOR YEAR LER NUMBER NUMBER

2 FOR REVISION NUMBER

8 TOTAL

7 2 FOR MONTH REPORT DATE

2 FOR DAY

4 FOR YEAR

UP TO 18 - FACILITY NAME

8 TOTAL - DOCKET NUMBER OTHER FACIUTIES INVOLVED

3 IN ADDITION TO 05000

9 OPERATING MODE

10 3 POWER LEVEL

I1 REQUIREMENTS OF 10 CFR

11 CHECK BOX THAT APPLIES

12 UP TO 50 FOR NAME MCENSEE CONTACT 14 FOR TELEPHONE

CAUSE VARIES

2 FOR SYSTEM

13 4 FOR COMPONENT EACH COMPONENT FAILURE

4 FOR MANUFACTURER

EPIX VARIES

14 14 CHECK BOX THAT 1 APPLIES SUPPLEMENTAL REPORT EXPECTED

8 TOTAL

15 2 FOR MONTH EXPECTED SUBMISSION DATE

2 FOR DAY

4 FOR YEAR

121 NUREG-1022, Rev. I
NRC FORM 368A U.S. NUCLEAR REGULATORY COMMISSWOr%

(4-951

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION

FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (31

YEAR I SEQUENTIAL I REVISION YER NUMBER iNUMBER

05000I OF

TEXT (If more space is required, use additional copies of NRC Form 366AJ 117)

NUREG-1022, Rev. 1 122

NRC FORM 366A (mm-yyyy)
NRC FORM 366B U.S. NUCLEAR REGULATORY COMMISSIOD

(MM-YYYY)

LICENSEE EVENT REPORT (LER)

FAILURE CONTINUATION

FACILITY NAME 11) II DOCKET II LER NUMBER 46) PAGE (3)

SEQUENTIAL REVISION Y NUMBER NUMBER

05000

I

OF

COMPIFTE ONE LINE FOR FACH COMPONENT AILURE I ) IN THIS REPORT 113) COMPLETE ONE LINE FOR EACH COMPONENT FAILURE I

CAUSE SYSTEM I COMPONENT MANUFACTURER REPORTABLE

TO EPIX CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE

I I TO EPIX

4. U 9 I t

a. a- a. I 4. 4 4.

9 9. 4 '9 9. 9* .9 9

4. I. .L I 1. 4. 4 4.

4. 9 I

-1. 4 4. 4

4. 4 9. .9

.4 1.

________________ I ______________ _________________ I E.~I 

_1-il__ 1 1 . _____________________ _________________

NRC FORM 3669 (MM-YYYY)

123 NUREG-1022, Rev. 1

APPENDIX A

HISTORICAL PERSPECTIVE ON EVENT REPORTING

NUREG-1022, Rev. I

Origin of 10 CFR 50.72 and 50.73

In December 1980, the U.S. Nuclear Regulatory Commission (NRC) determined that

requirements for reporting operational experience data needed major revision and approved the

development of an integrated operational experience reporting (IOER) system. The IOER

system was to combine, modify, and make mandatory the existing licensee event report (LER)

system and the industry supported, voluntary nuclear plant reliability data system (NPRDS).

The NPRDS contains both engineering and failure data submitted by nuclear power plant

licensees on specified plant components and systems. An advance notice of proposed

rulemaking concerning the IOER system was published on January 15, 1981 (46 FR 3541).

On June 8, 1981, the Institute of Nuclear Power Operations (INPO) stated it would assume

responsibility for managing and funding the NPRDS and would audit member utilities to assess

the adequacy of their participation in the NPRDS. The NRC believed the NPRDS would provide

the necessary operating experience data and further development of the IOER system was

discontinued.

On May 6, 1982, the NRC published a notice of proposed rulemaking in the Federal Register

(47 FR 19543) that would modify and codify the existing LER system. On July 26, 1983, after

consideration of public comments, the NRC published in the Federal Register (48 FR 33850) a

final rule under 10 CFR 50.73, which modified and codified the LER system and became

effective on January 1, 1984. In the rule, the Commission clearly indicated that the NPRDS is a

vital adjunct to 10 CFR 50.73 for component data.V) The purpose of the rule was to standardize

the reporting requirements for all nuclear power plant licensees, to eliminate reporting events of

low individual significance, and to require more thorough documentation and analyses of

reported events. Licensees are to submit such reports within 30 days of discovery. The

revised system also permits licensees to use the LER procedures for various other reports

required under specific sections of 10 CFR Part 20 and Part 50.

Also effective January 1, 1984, the NRC amended its immediate notification requirements of

significant events at operating nuclear power reactors (10 CFR 50.72) to clarify reporting criteria

and to require early reports only on those matters of value to the exercise of the Commission's

responsibilities. The amended rule was published in the Federal Register (48 FR 39039) on

August 29, 1983, and corrections to the rule (48 FR 40882) were published on September 12,

1983. Among the changes made were the use of terminology, phrasing, and reporting

thresholds similar to those of 10 CFR 50.73 whenever possible. Therefore, most events

reported under 10 CFR 50.72 also will require an in-depth followup report under 10 CFR 50.73.

NRC Workshops and Event Reporting Guidelines

In September 1983, the NRC staff published NUREG-1022, "Licensee Event Reporting

System," to provide supporting information and guidelines to persons responsible for the

) On January 1, 1997 NPRDS was replaced by a new reporting system entitled Equipment

Performance and Information Exchange (EPIX).

A-1 NUREG-1022, Rev. 1
preparation and review of LERs. NUREG-1022 includes (1) a brief description of how the NRC

analyzes LERs, (2) a restatement of the guidance contained in the Statements of Consideration

that accompanied the publication of the LER rule, (3) a set of examples of potentially reportable

events with staff comments on the actual reportability of each event, (4) guidelines on how to

prepare an LER and use the LER form, and (5) guidelines on submittal of LERs.

Between October 25 and November 16, 1983, the NRC held five regional workshops to discuss

the new LER rule (10 CFR 50.73) and the revised emergency notification rule (10 CFR 50.72).

Supplement I to NUREG-1022 was published in February 1984 to provide a summary of

answers to questions asked during the workshops.

Supplement 2 to NUREG-1022, issued in September 1985, contained evaluations of the quality

and completeness of an industry-wide sample of 415 LERs. The study was performed for the

NRC Office for Analysis and Evaluation of Operational Data (AEOD) by EG&G, Inc., at Idaho

National Engineering Laboratory. The report identifies deficiencies in LER content and

recommends corrective actions.

NRC Regulatory Imoact Study (Draft NUREG-1395)

In the fall of 1989, the NRC staff surveyed personnel from 13 nuclear power utilities to obtain

their views on the potential effect that NRC regulatory activities were having on the safe

operation of their nuclear plants. This survey was documented in NUREG-1395, "Industry

Perceptions of the Impact of the U.S. Nuclear Regulatory Commission on Nuclear Power Plant

Activities," Draft, March 1990. Section 8, "Reporting Events," of NUREG-1395 included

industry comments on reporting required by 10 CFR 50.72 and 50.73.

Specific industry concerns included the need for reporting

* inadvertent actuations of engineered safety feature (ESF) equipment

* actuation of ESF equipment involving no safety significance

* plant shutdowns required by plant technical specifications even though the action

statements of the technical specifications were being met

* grass fires not affecting plant safety

* radiation exposures in excess of regulatory limits

Amendment of 10 CFR 50.72 and 50.73

On September 10, 1992 the NRC published a final rule in the Federal Register (57 FR 41373)

to eliminate the requirement to report certain events which had been determined to be of little or

no safety significance. These were events that resulted in invalid actuation of several specific

engineered safety features.

NUREG-1022, Rev. I A-2
Other amendments have been made from time to time in order to make these sections conform

with changes in other sections. For example, the requirements for reporting releases of

radioactive material have been amended to conform with changes to 10 CFR Part 20.

Revision of NUREG-1022

Partially in response to the industry's concerns regarding event reporting described in NUREG1395, the NRC sponsored four additional regional workshops on event reporting during

September to November 1990.

The NRC staff determined that additional clarification was needed to further improve the

usefulness, quality, and threshold of reporting by the licensees under 10 CFR 50.72 and 50.73.

Therefore, a draft of Revision 1, to NUREG-1022, to encompass and supersede NUREG-1022

and Supplements I and 2, was published for comment in September 1991. In May 1992 and

again in May 1993, public meetings were held to discuss the issues raised by public comments.

After addressing these issues, a second draft was published for comment in February 1994.

The intent of this Revision 1 is to clarify reporting requirements of 10 CFR 50.72 and 50.73

without changing those reporting requirements. Accordingly, most of the guidance is not new or

different from generic reporting guidance previously published in the statements of

considerations for the rules, NUREG-1 022, its Supplements I and 2, or generic

correspondence such as generic letters and information notices. This final version of Revision

1 has been modified as appropriate to address public comments on the second draft that was

published for comment in February 1994. Further, at the staffs initiative the instructions for

preparation of LERs were augmented to address consistency of information provided in LERs

which is used to understand events, as discussed in Sections 2.8 and 5.

A-3 NUREG-1022, Rev. 1

APPENDIX B

EMERGENCY NOTIFICATION SYSTEM PROCESS

NUREG-1022, Rev. 1

NRC Prompt Response Personnel

Headquarters Operations Officer

The U.S. Nuclear Regulatory Commission (NRC) Operations Center is continuously staffed with

an NRC headquarters operations officer (HOO), who holds a degree in engineering and works

for the Office for Analysis and Evaluation of Operational Data (AEOD). HOOs are trained to

receive licensee notifications via the emergency notification system (ENS) made under Title 10

of the Code of Federal Regulations (10 CFR) Section 50.72. In addition, they are trained to

receive materials, security or transportation events, as well as inquiries from the public or

media. A second HOO is usually on duty during normal working hours to help with the more

frequent communications experienced during the work day.

Each HOO has previous nuclear experience and receives extensive classroom and simulator

training on both boiling-water and pressurized-water reactor systems at the NRC Technical

Training Center.

Although HOOs have a good general understanding of nuclear power plants, they do not have

expert knowledge of each specific plant. The HOOs ask questions and rely on the licensees to

explain plant-specific details, terms, and the limiting conditions for operation of related technical

specifications, to ensure they understand the significance of the event and are able to answer

pertinent questions. The HOOs will attempt to obtain all of the details of the event that bear on

its safety significance, even if those details would not otherwise be reportable.

The HOO determines, by procedure, how quickly the ENS event information needs to be

disseminated to various NRC officials and other Federal agencies and prepares a written report

of the oral ENS notification (ENS Event Notification Report) for electronic distribution to the

NRC Office of Nuclear Reactor Regulation (NRR), NRC regional offices and the Institute of

Nuclear Power Operations, by 7:30 a.m. each weekday morning.

Emergency Officer

If an emergency is declared or if it appears that the event may have significant plant-specific or

generic interest to the NRC, the HOO notifies the emergency officer (EO). The EO is assigned

on a weekly rotation from NRC staff members of the Senior Executive Service, and is on call 24

hours per day. These are typically NRR division directors, assistant division directors, or

branch chiefs, who are responsible for the NRC response to an event. The EO decides which

other NRC managers should be informed to participate in responding to the event. The EO

also participates in deciding whether the NRC Operations Center andlor the applicable NRC

regional incident response center will be partially or fully staffed to continuously monitor the

event.

Regional Duty Officer

The HOO promptly informs the regional duty officer (RDO) of any ENS notification affecting the

RDO's NRC region. The RDO, who is a senior NRC employee (typically a branch chief or

B-1 NUREG-1022, Rev. I
division director) in the applicable NRC region, is assigned a weekly rotation and is on call 24

hours per day. The RDO informs the responsible NRC section chief and other NRC staff, as

needed. The NRC regional staff follow up on the plant-specific aspects of each event through

the responsible section chief, resident inspectors, and other NRC managers or technical

experts, as needed.

Resident Inspector

If the safety significance of an event warrants or if the HOO can not obtain a clear

understanding of an event, the RDO may request a resident inspector to immediately

investigate, monitor, and report back to the NRC region and headquarters on the situation.

Licensees are encouraged to work with a resident inspector if they have a question regarding

the reportability of an issue. If the resident inspector cannot provide guidance, he or she can

direct the licensee through the region to headquarters for a more definitive discussion. The

resident inspector will not make the decision, but can advise what the regulations require. The

resident inspector should be informed about an event whenever an ENS notification is made.

The NRC relies on the continuously staffed NRC Operations Center, not the resident inspector,

to notify the appropriate NRC staff of a reportable event.

NRC Response to ENS Notifications

NRC Response Options

There is a wide range of typical NRC headquarters and region responses to an ENS

notification, depending on the safety significance of the event, including:

* The NRC Operations Center and the NRC regional incident response center may be

fully activated and a site team sent to the plant.

* Specific NRC staff may monitor the progress of the event from the NRC Operations

Center and/or regional incident response center and an NRC team may be sent to the

plant.

* A resident inspector may be requested to immediately investigate, monitor, and report

back to the NRC region and/or headquarters.

* Conference calls among NRC headquarters, region, and licensee management may be

established.

" The EO, RDO, and HOO may follow the progress of the event and request specific

information from the licensee on a periodic basis until the plant is in a safe condition.

* The RDO may receive the notification and contact the resident inspector for additional

information.

NUREG-1022, Rev. I B-2
Additional NRC Operating Event Review

Each working day the NRR Events Assessment and Generic Communications Branch (PECB)

and the AEOD Reactor Analysis Branch (RAB) obtain copies of notifications of events that were

received in the NRC Operations Center since the beginning of the previous working day.

Copies of the daily report from each regional office also are obtained. These reports present

the results of the regional offices' review of events occurring within the region since the

previous working day, regardless of whether licensees have submitted notifications under 10

CFR 50.72.

Each working day PECB and RAB personnel screen the notifications and regional daily reports

to identify events that are potentially significant. A telephone conference follows at a preset

time in the morning among representatives of PECB, RAB, the NRC Operations Center, and

others. The conference call is made to discuss the significance of the events and identify

specific events for further assessment. If an assessment is needed, engineers are assigned to

determine what happened during the event, what caused the event, what the consequences

might be, what corrective or preventive action is being taken, and whether that action is

sufficient. If the event is still ongoing, then the engineer follows its development.

During assessment of the event, the assigned engineer determines whether the event is

generic, significant, or both. The event is generic if other nuclear power plants have the

potential for occurrence of a similar event. Searches of plant operational experience data

bases may be performed by RAB personnel to identify similar occurrences and assess generic

applicability. The event is significant if any of the following occurred:

0 potential or actual degradation occurred in safety-related equipment or structures, fuel

integrity, the primary coolant pressure boundary, or containment

0 release of radioactivity (in excess of 10 CFR Part 20 limits) occurred

0 the plant was operated outside technical specification limits

* a scram with complications occurred

0 other conditions warranted attention by NRC

If the event is classified as significant, senior NRC management are informed at the next

weekly events briefing meeting. Briefing information, including event summaries and diagrams,

are placed in the Public Document Room (PDR). The event also is entered into the PECB

significant event tracking system. Each quarter the significant events are compiled and

published in the NRC performance indicator report ('Performance Indicators for Operating

Commercial Nuclear Power Reactors," issued by AEOD and available in the NRC PDR).

Additional event followup actions performed by NRR, the appropriate NRC regional office, and

AEOD personnel may include consulting with the Executive Director for Operations in the

selection of an incident investigation team (liT), participating in the decision to dispatch an

augmented inspection team (AiT) to the site and in the selection of the team members, or

performing a human performance evaluation at the plant. The appropriate NRC regional office

has the direct responsibility for routine followup and inspection related to reportable events.

B-3 NUREG-1022, Rev. 1
Depending on the number or types of event notifications by licensees, NRR also may issue

NRC generic letters, bulletins, and information notices.

NUREG-1022, Rev. 1 B-4
APPENDIX C

LICENSEE EVENT REPORT REVIEW PROGRAMS

NUREG-1022, Rev. I

Title 10 of the Code of Federal Regulations (10 CFR) Section 50.73 specifies that licensee

event reports (LERs) shall include a detailed narrative description of reportable operating

experience, including safety significant and potentially safety significant events and conditions.

By describing in detail the events or conditions required to be reported, LERs provide

information for detailed studies of events or conditions that might affect the health and safety of

the public.

Variations in LER counts from plant to plant can result from numerous factors, only one of

which is an actual difference in safety performance. Thus, the number of LERs submitted by a

plant should not be used as a measure of the plant's safety performance.

In addition to prompt followup to ENS notifications described in Appendix B, longer-term

followup of licensee events is conducted using the LER information. The appropriate U.S.

Nuclear Regulatory Commission (NRC) regional office conducts plant-specific followup, the

Office of Nuclear Reactor Regulation (NRR) conducts plant-specific and generic reviews, and

the Office for Analysis and Evaluation of Operational Data (AEOD) and its contracted national

laboratories, screen, classify, categorize, trend, assess, and store the data for each LER.

Those events and conditions, both plant-specific and generic, that appear to be important to

safety are further analyzed or evaluated. From this review process, the NRC determines

further actions such as (1) a special study initiated to propose revisions to regulatory programs,

(2) reporting as an abnormal occurrence to Congress, or (3) dissemination to the U. S. nuclear

power industry through generic communications and to the international community through the

Nuclear Energy Agency (NEA). The NEA is part of the Organization for Economic Cooperation

and Development and gathers information from its member countries on the operating

experience of commercial nuclear power plants worldwide.

Several fundamental objectives associated with the LER analysis process are

* to identify and quantify events and conditions that are precursors to potential severe

core damage

* to discover emerging trends or pattems of potential safety significance

0 to identify events that are important to safety and their associated safety concerns and

root causes and to determine the adequacy of corrective actions taken to address the

safety concerns

0 to assess the generic applicability of events

A precursor to potential severe core damage is an event or condition that could have been

serious if plant conditions, personnel action, or the extent of equipment failure or faulting had

been slightly different than that which occurred.

An analysis of trends and patterns in operational experience identifies repetitive events and

failures and searches past operating history for similar events and failures to determine if the

frequency of such events or failures is significant enough to be a cause for concern. When

C-1 NUREG-1022, Rev. I
appropriate, an NRC bulletin or information notice is issued or a generic study initiated to focus

on the nature, cause, consequences, and possible corrective actions of such a situation.

Trends and patterns analysis usually applies to events and conditions that individually are of

low safety significance but that become a safety significant factor because of repetition or, more

accurately, the frequency of occurrence.

AEOD studies of events that are important to safety are documented in the following reports:

* Case study reports document substantive, in-depth analyses of safety issues and the

bases for AEOD recommendations for regulatory or industry actions.

* Special study reports document accelerated assessments of significant operating events

and contain recommendations for remedial actions, if appropriate.

* Engineering evaluation reports document assessments of significant operating events

and contain suggestions for remedial actions, if appropriate.

* Technical review reports document studies of issues that were determined to have little

safety significance.

AEOD uses the sequence coding and search system (SCSS) data base for storage and

retrieval of LER data. This system, developed in the early 1980's and maintained under

contract at the Oak Ridge National Laboratory, Oak Ridge, Tennessee, contains an average of

150 items of information in its data base for each LER submitted since 1980.

AEOD uses LER data from the SCSS data base to support NRC activities such as plant

diagnostic evaluations, NRC senior management meetings, and performance indicators. The

SCSS data base also is a primary source of information for AEOD studies. In addition, NRC's

Office of Nuclear Regulation, Office of Nuclear Regulatory Research, and regional offices use

the SCSS as a source of information on operating experience.

AEOD also maintains LER information in the trends and patterns data base at Idaho National

Engineering and Environmental Laboratory (INEEL). This data base supports such specific

AEOD studies as those covering performance indicator data for reactor trips, safety system

actuations, and safety system failures. The INEL data base also is used to calculate forced

outage rates and equipment-forced outages per 1000 critical hours, as well as to support the

preparation of Commission site visit briefing packages, special studies, and the evaluations of

selected plants.

The information from LERs is widely used within the nuclear industry, both nationally and

internationally. For example, the industry's Institute of Nuclear Power Operation (INPO) uses

LERs as a basis for providing operational safety experience feedback data to individual utilities

through such documents as significant operating experience reports, significant event reports,

significant event notifications, and operations and maintenance reminders. U.S. vendors and

nuclear steam system suppliers, as well as other countries and international organizations, use

LER data as a source of operational experience data.

NUREG-1022, Rev. 1 C-2
APPENDIX D

10 CFR 50.72 INCLUDING

STATEMENT OF CONSIDERATIONS

Published in the Federal Register

On August 29, 1983

(Vol. 48, No. 168, pages 39039-39046)

This Federal Register notice does not provide a current version of 10 CFR 50.72,

which has been amended several times since 1983. Its purpose here is to

present the Statement of Considerations, which explains the basic reporting

requirements of 10 CFR 50.72.

NOTE:

NUREG-1022, Rev. 1

Federal Register / Vol. 48. No. 168 / Monday. Au2ust 29. 1983 / Rules and Reg'ulations a9039 • • . ... -- ... .. i l J I • _ _ • - II9- --

Commission. Washington. D.C. 20555:

Telephone (301) 492-4973.

SUPPLEMENTIARY INFORMATIOW.

I. Background

On February 29. 1980. the Commission

amended its regulations without prior

notice and comment to require timely

and accurate licensee reporting of

infornaition following significant events

at operating nuclear power reactors (45

FR 13434). The purpose of the rule was

to provide the Commission with

immediate reporting of twelve types of

significant events where immediate

Commission action to protect the public

health and safety may be required or

where the Commission needs accurate

and timely infornmation to respond to

heightened public concern. Although the

rule was made immediately effective.

comments were solicited. Many

corrmmenters believed the rule was in

some respects either vague and

ambiguous or overly broad.

After obtaining experience with

notifications required by the rule. the

Commission published in the Federal

Register a notice of proposed

rulemaking on December 21.1981 (46 FR

61894) and invited public comment. The

proposal was made to meet two

objectives: change 10 CFR 50.54 to

implement Section 201 of the NRCs 1980

Fiscal Year Authorization Act and

change 10 CFR 50.72 to more clearly

specify the significant events requiring

licensees to Immediately notify NRC.

The problems and issues which this

rulemaking addresses and the solutions

that it provides can be summarized in

five broad areas:

1. Authorization A ct for FW80

Section 201 of the Nuclear Regulatory

Commission Authorization Act for

Fiscal Year 1980 (Pub, L 96-295)

provides:

(a) Section 103 of the Atomic Energy Act of

1934 is amended by adding at the end thereof

the following new subsections: . Each license

Issued for a utilization facility under this

section or section 104b. shall require as a

condition thereof that in case of any accident

which could result in an unplanned release of

quantities of fission products In excess of

allowable limits for normal operation

established by the Commission. the licensee

shall Immediately so notify the Commission.

Violation of the condition prescribed by this

subsection may. in the Commission's

discretion, constitute grouinds for license

revocation. In accordance with section 187 of

this Act, the Commission shall promptly

amend each license for a utilization facility

Issued under this section or section 104b.

which is in effect on the date of enactment of

this subsection to Include the provisions

required under this subsection.

10 CFR Part 50

Immediate Notification Requirements

of Significant Events At Operating

Nuclear Power Reactors

AGENCr. Nuclear Regulatory

Commission.

ACTION. Final rule.

sUMMARV The Nuclear Regulatory

Commission is amending its regulations

which require timely and accurate

information from licensees following

significant events at commercial nuclear

power plants. Experience with existing

requirements and public comments on a

proposed revision of the rule indicate

that the existing regulation should be

amended to clarify reporting criteria and

to require early reports only on those

matters of value to the exercise of the

Commission's responsibilities. The

amended regulation will clarify the list

of reportable events and provide the

Commission with more useful reports

regarding the safety of operating nuclear

powerpants.

EFFECTIVE OATE: January 1.1984.

FOR FURTHER INFORMATION CONTACTI

Eric W. Weiss. Office of Inspection and

Enforcement. US. Nuclear Regulatory
39040 Federal Register / VoL 48. No. 168 / Monday. August 29. 1983 / Rules and Regulations

Accordingly. this rulemaking includes

an amendment to 10 CFR 50.54 that

would add an appropriate notification

requirement as a condition in the

operating license of each nuclear

utilization facility licensed under section

103 or 104b. of the Atomic Energy Act of

1954. as amended. 42 U.S.C. 2133. 2134b.

These facilities generally are the

commercial nuclear power facilities

which produce electricity for public

consumption. Research and test reactors

are not subject to the license condition

as they are licensed under section 104a.

or 104c. of the Act. Under the

amendment to 10 CFR 50.54. licensees

falling under sections 103 or 104b. would

be required. as a condition of their

respective operating licenses, to notify

the NRC immediately of events specified

in 10 CFR 502..

2 UnnecessoryReports

Several categories of reports required

by I 50.72 are not useful to the NRC.

Among these categories are reports of:

worker injury, small radioactive

releases, and minor security problems.

For exam ple. reports are presently

required iW a worker onsite experience

chest pains or another illness not related

to radiation and is sent to a hospital for

evaluation: or if the vent stack monitor

moves upward a few percent yet

radiation levels remain 100MO times

below technical specification limits; or if

the security computer malfunctions for a

few minutes.

This rulemaking eliminates such

reporting requirements from I 50.72 and

in general clarifies and narrows the

scope of reporting. However. revision of

Part 73 of the Commission's regulations

is necessary to resolve all problems with

security reports.

3. Terminology. Phrasing and Reporting

Thresholds

The various sections of 10 CFR 50

have different phrasing. terminology.

and thresholds in the reporting criteria.

Even when no different meaning is

intended a change in wording can cause

confusion.

This rulemaking has been carefully

written to use te-nindlagy. phrasing.

and reporting thresholds that are either

identical to or similar to those in I 50.73.

whenever possible. Other conforming

amendments to Parta 20. 1.73. and in

150.55 and Appendix E of Part 50 are

under development.

As a parallel activity to the

preparation of) 50. on July 2. 1983.

the Commission has published a

Licnse Event Report WER) Rule

( 50.73) which requires licensees for

operating nuclearpower planls to

prepare detailed writtem reports for

certain events (48 Fr 33850).

4. Coordination with Licensee's

Emergency Plan

The current scheme for licensees*

emergency plans includes four

Emergency Classes. When the licensee

declares one of the four Emergency

Classes. it must report this to the

Commission as required by 1 50.72 The

lowest of the four Emergency Classes.

Notification of Unusual Event. has

resulted in unnecessary emergency

declarations. Events that fall within the

Unusual Event class have been neither

emergencies in themselves nor

precursors of more serious events that

are emergencies.

Although changes to the definition of

the Emergency Classes are not being

made in this rulemaking, a new

reporting scheme that would ultimately

eliminate "Unusual Event" as an

Emergency Class requiring notification

can be adopted consistent with this rule.

A proposed rulemaking which would

redefine the Emergency Classes in

150.47 Is in preparation and may soon

be published for public comment. This

final rulemaking makes possible the

elimination of "Unusual Event" as an

emergency class without further

amendment of I 50.72 by including in

the category of non-Emergencies the

subcategory of "one-hour reports."

& Vogue or Ambiguous Reporting

Criteria

The reporting criteria in 1 50.72 have

been revised in order to clarify their

scope and intent. The criteria were

revised for the proposed rule and in

response to public comment. The

"Analysis of Comments" portion of this

Fideral Register notice describes In

more detail specific examples of

changes In wording Intended to

eliminate vagueness or ambiguity.

IL Analysis of Comments

Twenty letters of comment were

received in response to the Federal

Register notice published on December

21.1981 (48FR 81894).' Of the twenty

letters of comment received. the vast

majority 123 of 203 were from utilities

owning or operating nuclear power

plants. This Federal Register notice

described the proposed revision of 10

CFR 507Z 'Notificatlon of Significant

Events." and 10 CFR 50.%54 "Conditions

of Licenses." A discussion of the more

significant conatnents follows:

I Copf of thea. documeta an available f

publk inspecting and copyhn fr a fet Is the 7•C

Public Dwmnteat Room. IM5 H Stmnt. N.W.

Wasotlm.=D..5

Conditions of Licenses (§f50.54

A few commenters said that the

"Commission already has the ability to

enforce its regulations and does not

need to incorporate the items as now

proposed into conditions of license."

The Commission has decided to

promulgate the proposed revision of

§ 50.54. -Conditions of Licenses." in

order to satisfy the intent of Congress as

expressed in Section 201 of the Nuclear

Regulatory Commission Authorization

Act for Fiscal Year 1981. This Act and

its relationship to J 50.54 are discussed

in detail in the Fedcral Register notice

for the proposed rule 146 FR 61894).

Coordination With Other Reporting

Requirements (Final Rule §f5O.72)

Seven commenters said that the NRC

should coordinate the requirements of 10

CFR 50.72 with other rules. with

NUREG-0CX. "Criteria for Preparation

and Evaluation of Radiological

Emergency Response Plans and

Preparedness in Support of Nuclear

Plants." and with Regulatory Guide 1.18.

"Reporting of Operating Information

." Many of these letters Identified

overlap. duplication, and inconsistency

among NRC's reporting requirements.

The Commission is making a

concerted effort to ensure consistent

and coordinated reporting requirements.

The requirements contained in the

revision of 10 CFR 50.72 are being

coordinated with revision of 1 5..

I 50.55(e). Appendix E of Part 50.

120.402. 1 73.71. and Part 21.

Citing ZO CFR 50.72 as a Basis for

Notification [Final Rule $.a72(a)(4)]

A few commenter objected to citing

I 50.72 as a basis when making a

telephone notification. The letters of

comment questioned the purpose. legal

effect. and burden on the licensee.

The Commission does not believe that

It Is an unnecessary burden for a

licensee to know and Identify the basis

for a telephone notification required by

I 50.72. There have been many

occasions when a licensee could not fell

the NRC whether the telephone

notification was beilng made in

accordance with Technical

Specifications. 10 CFR 50.72 some other

requirement. or was just a courtesy call.

Unless the licensee can Identify the

nature of the report, It ti difficult for the

NRC to know what significance the

licensee attaches to the report. and it

becomes more difficult for the NRC to

respond quickly and properly to the

event.
Federal Register I Vbl. 48. No. 168 / Monday. August 29. 1983 / Rules and Regulations 39041

WON I

Immediate Shutdown (Final Rule

§ 50.72(b)(j)(i))

Several commenters objected to the

use of the term. "immediate shutdown."

saying that Technicai Specifications do

not use such a term.

The term is used In some but not all

Technical Specifications. Consequently.

the Commission has revised the

reporting criterion in question. The final

rule requires a report upon the initiation

of any nuclear power plant shutdown

required by Technical Specifications.

Plant Operating end.Emergency

Procedures (Final Rule I 5072(b)(1J(ii))

Several commenters said that the

reporting criteria should not make

reference to plant operating and

emergency procedures because:

a. It would take operators too long to

decide whether a plant condition was

covered by the procedures.

b. The procedures cover events that

are not of concern to the NRC. and

c. The procedures vary from plant to

plant.

While the plant operating personnel

should be familiar with plant

procedures. it is true that procedures

vary from plant to plant and cover

events other than those which

compromise plant safety. However. the

wording of the reporting criteria has

been modified (I 50.72(b)(l}[ii) in the

final rule) to narrow the reportable

events to those that significantly

compromise plant safety.

Notwithstanding the fact that the

procedures vary from plant to plant. the

Commission has found that this criterion

results in notifications indicative of

serious events. The narrower, more

specific wording will make it possible

for plant operating personnel to identify

reportable events under their specific

operating procedures.

Building Evacuation (Final Rule

I 5sa.2 7(b)(iii))

Ten commenters said that the

proposed I 50.72(b)(6)[il) regarding

"any accidental, unplanned or

uncontrolled release resulting in

evacuation of a building" was unclear

and counterproductive in that it could

cause reluctance to evacuate a building.

Many of these commenters stated that

the reporting of in-plant releases of

radioactivity that require evacuation of

individual rooms was Inconsistent with

the general thrust of the rule to require

reporting of significant events. They

noted that minor spills, small gaseous

waste releases, or the dipturbance of

contaminated particulate matter (e.g.

dust) may all require the temporary

evacuation of Individual rooms until the

airborne concentrations decrease or

until respiratory protection devices are

utilized. They noted that these events

are fairly common and should not be

reportable unless the required

evacuation affects the entire facility or a

major part of it.

The Commission agrees. The wording

of this criterion has been changed to

include only those events which

significantly hamper the ability of site

personnel in performance of duties

necessary for safe operation.

One commenter was concerned that

events occurring on land owned by the

utility adjacent to its plant might be

reportable. This is not the intent of this

reporting requirement. The NRC is

concerned with the safely of plant and

personnel on the utility's site and not

with non-nuclear activities on land

adjacent to the plant.

Explicit 7hreats (Final Rule

S50.72(b)(1)(vi))

A few commenters said that the intent

of the term. "explicitly threatens." was

unclear. Those commenting wondered

what level of threat was involved. The

term. "explicitly threatens. "has been

deleted from the final rule. Instead, the

final rule refers to "any event that poses

an actual threat to the safety of the

nuclear power plant"Ji 50.72(b)(1)(vi)]

and gives examples so that it is clear the

Commission Is interested in real or

actual threats as opposed to threats

without credibility.

Nobfication Timing (Final Rule

if50.74b)(2))

The commenters generally had two

points to make regarding the timing of

reports to the NRC. First. the comments

supported notification of the NRC after

appropriate State or local agencie'have

been notified. Sccond. two commenters

requested a new four-to six-hour report

category for events not warranting a

report with one hour.

Allowing more time for reporting

some non-Emergency events would

lessen the impact of reporting on the

individuals responsible for maintaining

the plant in a safe condition. Limiting

the extension of the deadline to four

hours ensures that the report is made

when the information is fresh in the

minds of those involved and that It is

more likely to be made by those

Involved rather than by others on a later

shift.

Other. more significant nonEmergency eventu and all declarations

of an Emergency must continue to be

reported within one hour. The one-hour

deadline is necessary If the Commission

is to fulfill Its responsibilities during and

following the most serious events

occurring at operating nuclear power

plants. A deadline shorter than one hour

was not adopted because the

Commission does not want to interfere

with the operator's ability to deal with

an accident or transient in the first few

critical minutes.

Thcrefore. based on these comnments

and its experience the NRC has

established a "four-hour report." as was

suggested.

Reactor Scrams (Final Rule

§ 50.72[bJt2Jfii))

Several commenters said that reactor

scrams, particularly those scrams below

power operation, should not require

notification of the NRC within one hour.

In response to these comments, the

Commission had changed the reporting

deadline to four hours. However. the

Commission does not regard reactor

scrams as "non-events." as stated in

some letters of comment. Information

related to reactor scrams has been

useful in identifying safety-related

problems. The Commission agrees that

four hours is an appropriate deadline for

this reporting requirement because these

events are not as important to

immediate safety as are some other

events.

Radioactive Release Threshold (Final

Rule § 50.72(b)(2)(iv))

Several commenters said that the

threshold of 25% of allowable limits for

radioactive releases was too low for

one-hour reporting.

Based upon these comments and its

experience. the Commission has

changed the threshold of reporting to

those releases exceeding two times Part

20 concentrations when averaged over a

period of one hour. This will eliminate

reports of releases that represent

negligible risk to the public.

The Commission has found that low

level radioactive releases below two

times Part 20 concentrations do not. in

themselves, warrant immediate

radiological response.

This paragraph requires the reporting

of those events that cause an unplanned

or uncontrolled release of a significant

amount of radioactive material to offsIte

areas. Unplanned releases should occur

infrequently; however, when they occur.

at least moderate defects have occurred

in the safety design or operational

control established to avoid their

occurrence and. therefore. these events

should be reported.

Personnel Radioactive Contamination

(Final Rule I 50,,2(bX2Xv))

Several commenters objected to the

use of vague terms such as "extensive
39042 Federal Register / Vol. 48, No. 108 / Monday, August 29, 1983 / Rules and Rctulations_

onsite contamination" and "readily

removed" in one of the reporting criteria

of the proposed rule.

Based on this comment, new criteria

have been prepared that use more

specific terms. For example, one new

criterion requires reporting of "Any

event requiring the transport of a

radioactively contaminated person to an

offsite medical facility for treatment."

Experience with telephone notifications

made to the NRC Operations Center

suggests that this new criterion will be

easily understood.

Ill. Paragraph-by-Paragraph Explanation

of the Rule

Paragraph 50.72=() reflects some

consolidation of language that was

repeated in various subparagraphs of

the proposed rule. In general. the intent

and scope of this paragraph do not

reflect any change from the proposed

rule.

Several titles were added to this and

subsequent sections. For example.

paragraph 50.72(b) Is titled "NonEmergency Events" and it has two

subparagraphs. (b)(1). titled. "One-Hour

Reports" and (b)(2), "Four-Hour

Reports." The events which have a onehour deadline are those having the

potential to escalate to an Emergency

Class. The four-hour deadline is

explained in the analysis of paragraph

(b)(2).

Paragraph ,Q72(b),IXi(A) requires

reporting of "The initiation of any nuclear plant shutdown required by

Technical Specifications." Although the

intent and scope have not changed. the

change in wording between the

proposed and final rule Is intended to

clarify that prompt notification is

required once a shutdown is initiated.

In response to public comment. the

term "immediate shutdown" that was

used in the proposed rule is not used in

the final rule. The term was vague and

unfamiliar to those licensees who did

not have Technical Specifications using

the term.

This reporting requirement is intended

to capture those events for which

Technical'Specifications require the

initiation of reactor shutdown. This will

provide the NRC with early warning of

safety significant conditions serious

enough to warrant shutdown of the

plant.

Pm'ogrph 50 )(bX1X(i)(B) was added

to be consistent with existing

requirements in 50.54(x) and the

existing I 50.2c) as published in the

Federal Register on April 1983 (48 FR

13966) which require the licensee to

notify the NRC Operations Center by telephone when the licensee departs

from a license condition or technical

specification.

Paragraph 50.72(b)(l)(ii).

encompassing events previously

classified as Unusual Events and some

events captured by proposed

§ 50.72(b)(1) was added to provide for

consistent, coordinated reporting

requirements between this rule and 10

CIF-R 50.73 which has a similar provision.

Public comment suggested that there

should be similarity of terminology.

phrasing. and reporting thresholds

between i 50.72 and 1 50.73. The intent

of this paragraph is to capture those

events where the plant. including its

principal safety barriers, was seriously

degraded or in an unanalyzed condition.

For example. small voids in systems

designed to remove heat from the

reactor core which have been previously

shown through analysis not to be safety

significant need not be reported.

However. the accumulation of voids that

could inhibit the ability to adequately

remove heat from the reactor core,

particularly under natural circulation

conditions, would constitute an

unanalyzed condition and would be

reportable. In addition, voiding In

instrument lines that results In an

erroneous indication causing the

operator to misunderstand the true

condition of the plant is also an

unanalyzed condition and should be

reported.

The Commission recognizes that the

licensee may use engineering judgment

and experience to determine whether an

unanalyzed condition existed. It Is not

intended that this paragraph apply to

minor variations in individual

parameters, or to problems concerning

single pieces of equipmentL For example.

at any time. one or more safety-related

components may be out of service due

to testing. maintenance. or a fault that

has not yet been repaired. Any trivial

single failure or minor error in

performing surveillance tests could

produce a situation in which two or

more often unrelated, safety-grade

components are out-of-service.

Technically. this is an unanalyzed

condition. However. these events should

be reported only if they involve

functionally related components or If

they significantly compromise plant

safety. When applying engineering

judgement. and there is a doubt

regarding whether to report or not. the

Commission's policy is that licensees

should make the report.

Finally. this paragraph also Includes

material (e.g.. metallurgical or chemical)

problems that cause abnormal

degradation of the principal safety

barriers (Le., the fuel cladding, reactor

coolant system pressure boundary, or

the containment). Examples of this type

of situation include:

(a) Fuel cladding failures in the

reactor, or in the storage pool. that

exceed expected values, or that are

unique or widespread. or that are

caused by unexpected factors, and

would involve a release of significant

quantities of fission products.

(b) Cracks and breaks in the piping or

reactor vessel (steel or prestressed

concrete) or major components in the

primary coolant circuit that have safety

relevance (steam generators. reactor

coolant pumps. valves, etc.).

(c) Significant welding or material

defects in the primary coolant system.

(d) Serious temperature or pressure

transients.

(e) Loss of relief and/or safety valve

functions during operation.

(f) Loss of containment function or

integrity including:

(i) Containment leakage rates

exceeding the authorized limits,

(ii) Loss of containment isolation

valve function during tests or operation.

(iii) Loss of main steam isolation

valve function during test or operation.

or

(iv) Loss of containment cooling

capability.

Paragraph 5072(b)[()(iii)

encompassing a portion of proposed

50.72(b)(2). was reworded to correspond

to a similar provision of 10 CFR

50.73(a)(2](iii). Making the requirements

of 10 CFR 50.72 and 50.73 similar in

language increases the clarity of these

rules and minimizes confusion.

The paragraph has also been

reworded to make it clear that It applies

only to acts of nature (e.g., tornadoes)

and external hazards (e.g.. railroad tank

car explosion). References to acts of

sabotage have been removed, since

these are covered by 1 73.71. In addition.

threats to personnel from Internal

hazards (e.gq. radioactivity releases) that

hamper personnel in the performance of

necessary dulies are now covered by

paragraph ,M72(b)(1)(vi). This paragraph

covers those events involving an actual

threat to the plant from an external

condition or natural phenomenon. and

where the threat or damage challenges

the ability of the plant to continue to

operate in a safe manner (including the

orderly shutdown and maintenance of

shutdown conditions). The licensee

should decide if a phenomenon or

condition actually threatens the plant.

For example. a minor brush fire In a

remote area of the site that Is quickly

controlled by fire fighting personnel and.

as a result, did not present P t!hreat to

the plant should not be reported.

However. a major forest fire, large-scale
Federal Register / Vol. 48. No. 108 / Monday. August 29. 1983 / Rules and Regulations 39043

flood, or major earthquake that presents

a clear threat to the plant should be*

reported. As another example, an

Industrial or transportation accident

which occurs near the site. creating a

plant safety concern, should be

reported.

Paragraph .50.7itb/i)(iv).

encompassing events previously

classified as Unusual Events. requires

the reporting of those events that result

in either automatic or manual actuation

of the ECCS or would have resulted in

activation of the ECCS if some

component had not failed or an operator

action had not been taken.

For example. if a valid ECCS signal

were generated by plant conditions, and

the operator were to put all ECCS

pumps in pull-to-lock, though no ECCS

discharge occurred, the event would be

reportable.

A "valid signalr refers to the actual

plant conditions or parameters

satisfying the requirements for ECCS

initiation. Excluded from this reporting

requirement would be those instances

where Instrument drift, spurious signals.

human error, or other invalid signals

caused actuation of the ECCS. However.

such events may be reportable under

other sections of the Commission's

regulations based upon other details: In

particular, paragraph 50.72(b)(2)(ii)

requires a report within four hours if an

Engineered Safety Feature (ESF) Is

actuated.

Experience with notifications made

pursuant to J 50.72 has shown that

events involving ECCS discharge to the

vessel are generally more serious than

ESF actuations without discharge to the

vessel. Based on this experience, the

Commission has made this reporting

criterion a "One-Hour Report." Parograph 50. 7b])(b1v).

encompassing events previously

classified as Unusual Events. covers

those events that would impair a

licensee's ability to deal with an

accident or emergency. Notifying the

NRC of these events may permit the

NRC to take some compensating

measures and to more completely assess

the consequences of such a loss should

it occur. during an accident or

emergency.

Examples of events that this criterion

is intended to cover are those in which

any of the following are not available:

1. Safety parameter display system

(SpDS).

2. Emergency Response Facilities

(ERFs).

3. Emergency communications

facilities and equipment Including the

Emergency Notification system (ENS).

4. Public prompt Notification System

including sirens.

S. Plant monitors necessary for

accident assessment.

Paraogrph 5O.72'bffI)(vi).

encompassing some portions or the

proposed §I 50.72(b) (2) and (6). has

been revised to add the phrase.

"including fires, toxic gas releases, or

radioactive releases." This addition

covers the "evacuation" portion of

paragraph 50.72(b)06)(iii) of the proposed

rule. This change in wording for the final

rule was made in response to public

comments discussed above.

While paiagraph 50.72(b)[1}{iii) of the

final nile primarily captures acts of

nature, paragraph 50.72(b)(1)(vi)

captures other events, particularly acts

by personnel. The Commission believes

this arrangement of the reporting criteria

in the final rule lends Itself to more

precise interpretion and is consistent

with those pubic comments that

requested closer coordination between

the reporting requirements In this rule

and other portions of the Commission's

regulations.

This provision requires reporting of

events, particularly those caused by acts

of personnel, which endanger the safety

of the plant or interfere with personnel

in performance of duties necessary for

safe plant operations.

The licensee must exercise some

Judgment in reporting under this section.

For example, a small fire on site that did

not endanger any plant equipment and

that did not and could not reasonably be

expected to endanger the plant, is not

reportable.

Paragraph 50.721b)(1) of the proposed

rule was split into § 50.72[b)f()ii) and

§50..72(b()2)(i) in the final rule In order

to permit some type of reports to be

made within four hours instead of one

hour because these reports have less

safety significance. In terms of thet'r

combined effect, the overall Intent and

scope of these paragraphs have not

changed from those in the proposed rule.

Since the types of events intended to be

captured by this reporting requirement

are similar to I 5O.72(b)(1)(li), except

that the reactor Is shut down. the reader

should refer to the explanation of

I 50.72(b)(1)(ii) for more details on

intent.

Palagraph 5.72(b](2) Although the

reporting criteria contained in the

subparagraphs of I 50.72(b)(2) were in

the proposed rule. in response to public

comment the Commission established

this "Non-Emergency" category for

those events with slightly less urgency

and less safety significance that may be

reported within four hours Instead of

one hour.

The Commission wants to obtain such

reports from personnel who were on

shift at the time of the event, when this

is pu.sihle. because these personnel will

have a better knowledge of the

circumstances associated with the vent.

Reports made within four hours of the

event %.hould make this possible while

not imposing the more rigid one hour

requirements.

The• ,porting requifemnnt in

puryn.-:ph 50.721bl121(i) is similar to a

requiirement it J 50.73. Moreover. exrtcpt

for w•f.rring Its a shutdown reactor, this

reporting requirement is also similar to

the "One-I hlur Report" in

J 50.72:1oll,)(1](. However this paranraph

applies to a reactor in shutdown

condition. Fvents within this

requirement have less ur.cncy and can

be reported within four hours as a "NonEmergency."

Paragraph 5.(72(b)(2)(ii) (proposed

50.7211)b5)) Is made a "Non-Emergency"

in response to public comment, because

the Commission agrees that the covered

events generally have slightly less

urgency and safety significance than

those events included In the "One-I lour

Reports."

The Intent and scope of this reporting

requirement have not changed from the

proposed rule. This paragraph Is

intended to capture events during which

an ESF actuates, either manually or

automatically, or falls to actuate. ESF&

are provided to mitigate the

consequences of the event: therefore. (I]

they should work properly when called

upon and (2) they should not be

challenged unnecessarily. The

Commission is Interested both in events

where an ESF was needed to mitigate

the consequences of the event (whether

or not the equipment performed

properly) and events where an ESF

opcntted unnecessarily.

"Actuation" of multichannel ESF

Actuation Systems is defined as

actuation of enough channels to

complete the minimum actuation logic.

Therefore. single channel actuations.

whether caused by failures or otherwise,

are not reportable If they do not

complete the minimum actuation logic.

Operation or an E$F as part of a

planned test or operational

evolution need not be reported.

I lowevcr, if during the test or

evolution the ESF actuates in a way that

is not part of the planned procedure,

that actuation should be reported. For

exampld. if the normal reactor shutdown

procedure requires that the control rods

be inserted by a manual reactor trip, the

reactor trip need not be reported.

However, If conditions develop during

the shutdown that require an automatic

reactor trip, such a reactor trip should

be reported. The fact that the safety
9044 Federal Register / Vol. 48. No. 168 / Monday. August 29. 1983 1 Rules and Regulations

nalysis assumes that an ESF will

ctuate automatically during an event

oes not eliminate the need to report iat actuation. Actuations that need not

e reported are those initiated for

easons other then to mitigate the

onsequences of an event (e.g.. at the

iscretion of the licensee as part of a

lanned procedure).

,Parogroph 50.72(b)(2)(iii) (proposed

O.721b)(4)) has been revised and

,implified.

The words "any instance of personal

.ror. equipment failure, or discovery of

lesign or-procedural inadequacies" that

ippeared in the proposed rule have been

eplaced by the words "event or

:ondition." This simplification in

anguage is Intended to clarify what was

i confusing phrase to many of those

vho commented on the proposed rule.

'Uso in response to public comment, this

eportlng requirement is a "Nonmergency" to be reported within four

iours instead of within one hour.

This paragraph is based on the

issumption that safety-related systems and structutres intended to mitigate

the consequences of an accident. While

paragraph 50.72(b](2)(ii) applies to

actual demands for actuation of an ESF.

paragraph 50.72(b)(2)(iii) covers an

event where a safety system could have

failed to perform its intended functi6n

because of one or more personnel errors,

including procedure violations;

eaqipment failures; or design, analysis, rication, construction, or procedural deficiencies. The event should be

reported regardless of the situation or

condition that caused the structure or

system to be unavailable.

This reporting requirement is similar

to one contained in I 0.3 thus

reflecting public comment identifying the need for closer coordination of

reporting requirements between 1 50.72

and 1 50.73.

This paragraph includes those safety

systems designed to mitigate the

consequences of an accident (e.g., containment Isolation. emergency

filtration). Hence. minor operational

events such as valve packing leaks,

which could be considered a lack of

control of radioactive material, should

not be reported under this paragraph.

System leaks or other similar events

may. however, be reportable under other

paragraphs.

This paragraph does not include those

cases where a system or component is

removed from service as part of a

planned evolution, in accordance with

an apptoved procedure, and in

accordance with the plant's Technical

Specifications. For example, If the

licensee removes part of a system from

service to perform maintenance, and the

Technical Specifications permit the

resulting configuration. and the system

or component is returned to service

within the time limit specified In the

Technical Specifications. the action

need not be reported under this

paragraph. However. if, while the

component is out of service, the licensee

identifies a condition that could have

prevented the system from performing

its intended function (e.g.. the licensee

finds a set of relays that is wired

incorrectly), that condition must be

reported.

It should be noted that there are a

limited number of single-train systems that perform safety functions (e.g.. the

High Pressure Coolant Injection System in BWRs). For such systems, loss of the

single train would prevent the

fulfillment of the safety function of that

system and. therefore, must be reported

even though the plant Technical

Specifications may allow such a

condition to exist for a specified length

of time. Also, if a potentially serious

human error is made that could have

prevented fulfillment of a safety

function. but recovery factors resulted In

the error being corrected, the error is

still reportable. The Commission recognizes that the

application of this arid other paragraphs

of this section involves a technical

judgment by licensees. In this case, a

technical judgment must be made

whether a failure or operator action that

disabled one train of a safety systemcould have, but did not. affect a

redundant tr*in. If so. this would

constitute an event that "could have

prevented" the fulfillment of a safety

function. and. accordingly, must be

reported.*

If a component fails by an apparently

random mechanism, it may or may not

be reportable if the functionally redundant component could fail by the

same mechanism. To be reportable, it is

necessary that the failure constitute a

condition where there Is reasonable

doubt that the functionally redundant

train or channel would remain

operational until It completed Its safety function or is repaired. For example, if a

pump fails because of Improper

lubrication. there Is a reasonable

expectation that the functionally

redundant pump. which was also

improperly lubricated, would have also

failed before it completed Its safety

function. then the failure Is gportable

and the potential failure of the

functionally redundant pump mukt be

reported.

Interaction between systems.

particularly a safety system and a nonsafety system, Is also included in this

criterion. For example. the Commission

is increasingly concerned about the

effect of a loss or degradation of what

had been assumed to be nonessential

inputs to safety systems. Therefore. this

paragraph also includes those cases

where a service (e.g.. heating.

'ventilation. and cooling) or input (e.g..

compressed air) which is necessary for

reliable or long-term operation of a

safety system is lost or degraded. Such

loss or degradation is reportable. if the

proper fulfillment of the safety function

is not or can not be assured. Failures

that affect inputs or services to systems that have no safety function need not be

reported.

Finally. the Commission recognizes

that the licensee has to decide when

personnel actions could have prevented

fulfillment of a safety function. For

example, when an Individual Improperly

operates or maintains a component. that

person might conceivably have made

the same error for all of the functionally

redundant components (e.g.. if an

individual incorrectly calibrates one

bistable amplifier in the Reactor

Protection System. that person could

conceivably incorrectly calibrate all

bistable amplifiers). However. for an

event to be reportable it is necessary

that the actions actually affect or

involve components in more than one

train or channel of a safety system, and

the result of the actions must be

undersirable from the perspective of

protecting tho health and safety of the

public. The components can be

functionally redundant (e.g., two pumps

in different trains) or not functionally

redundant (e.g.. the operator correctly

stops a pump in Train "A" and.'instead

of shutting the pump discharge valve in

Train "A," he mistakenly shuts the

pump discharge valve in Train "rn.

Pabragrphs 5a72(b)2Xi7) (proposed 50.72(b)(6]) has been changed to clarify

the requirement to report releases of

radioactive material. The paragraph is

similar to I 20.403 but places a lower

threshold for reporting events at

commercial power reactors. The lower

threshold Is based on the significance of

the breakdown of the licensees program

necessary to have a release of this size.

rather than on the significance of the

impact of the actual release. The

existing licensee radioactive material

effluent release monitoring programs and their associated assessment

capabilities are sufficient to satisfy the

intent of 50.72(b](2](iv).

Based upon public comment and a

reevaluation by the Commission staff.

the reporting threshold has been

changed from "25%" in the proposed rule

to "2 times" in the final rule and has
Federal Register / Vol. 48. No. 168 / Monday. August 29. 1983 / Rules und ReM.lations 39045

been reclassified as a "Non-Emergency"

to be reported within four hours initead

of within I hour.

Also this reporting requirement has

been changed to make a more uniform

requirement by referring to specific

release criteria instead of referring only

to Technical Specifications that may

var- somewhat among facilities.

this reporting requirement is intended

to capture those events that may lead to

an accident situation %here significant

amounts of radioactive material could

be released from the facility. Unplanned

releases should occur infrequently:

however, if they occur at the levels

specified. at least moderate detects have

occurred in the safety design or

operational control established to avoid

their occurrence and. therefore, such

events should be reported.

Normal operating limits for

radioactive effluentreleases are based

on the limits of 10 CFR Part 20 which

establishes maximum annual average

concentration in unrestricted areas. This

reporting requirement addresses

concentrations averaged over a one

hour period and represents less than

0.1% of the annual quantities of

radioactive materials permitted to be

released by 10 CFR Part 20.

Paragraph 50.2[b)(2)fv] (proposed

rule 50.72(b)(71) has three changes. The

first eliminates the phrase "occurring

onsite" because it Is implied by the

scope of the rule. The second replaces

"injury involving radiation" with "radioactively contaminated person."

This change was made because of the

difficulty in defining injury due to

radiation, and more Importantly.

because 10 CFR Part 20 captures events

involving radiation exposure.

The third change. In response to

public comment, was to make this

reporting requirement a four-hour

notification. instead of one-hour

notification. This change was made

because these events have slightly less

safety significance than those required

to be reported within one hour.

Paragraph 50.?2[b)(2)(vi) (not in

proposed rule) besides coverjng some

events such as release of radioactively

contaminated tools or equipment to the

public that may warrent NRC attention.

also covens those events that would not

otherwise warrant NRC attention except

for the interestof the news media, other

government agencies or the public. In

terms of its effect on licensees, this is

not a new reporting requirement

because the threshold for reporting

injuries and radioactive release was

much lower under the proposed rule.

This criterion will capture those events

previously reported under other criteria

when such events require the NRC to

respond because of media or public

attention.

Paragraph .50.72(c) (proposed 50.72(c))

has remained essentially unchanged

from the proposed rule. except for

addition of the title "Followup

Notification" and some renumbering.

This paragraph is intended to provide

the NRC with timely notification when

an event becomes more serious or

additional information or new analyses

clarify an event.

This paragraph also permits the NRC

to maintain a continuous

communications channel because of the

need for continuing follow-up

information or because of

telecommunications problems.

IV. Regulatory Analysis

Th7 Commission has prepared a

regulatory analysis on this regulation.

The analysis examines the costs and

benefits of the Rule as considered by the

Commission. A copy of the regulatory

analysis is available for inspection and

copying for a fee at the NRC Public

Document Room. 1717 H Street. NW..

Washington. D.C. Single copies of the

analysis may be obtained from Eric W.

Weiss. Office of Inspection and.

Enforcement. U.S. Nuclear Regulatory

Commission. Washington. D.C. 20555.

Telephone (301) 492-4973.

V. Paperwork Reduction Act Statement

The information collection

tequirements contained In this final rule

have been approved by the Office of

Management and Budget pursuant to the

Paperwork Reduction Act. Pub. L 96-511

(clearance number 3150-0011).

VL Regulatory Flexibility Certification

"In accordance with the Regulatory

Flexibility Act of 1980. U.S.C. 605(b)

the Commission hereby certifies that

this regulation will not have a

significant economic impact on a

substantial number of small entities.

This final rule affects electric utilities

that are dominant in their respective

service areas and that otvn and operate

nuclear utilization facilities licensed

under sections 103 and 104b. of the

Atomic Energy Act of 1954. as amended.

The amendments clarify and modify

presently existing notification

requirements. Accordingly. there is no

new. significant economic impact on

these licensees, nor do the affected

licensees fall within the scope of the

definition of "sniall entities" set forth in

the Regulatory Flexibility Act or within

the Small Business Size Standards set

forth in regulations issued by the Small

Business Administration at 1-L CFR Part

121.

List uf Subjects in 10 CFR Part 50

Antitrust. Classifie'd Information. Fire

prevention. Incorporation by reference.

Interguvernmental relations. Nuclear

power plants and reactors. Penalty.

Radiailion protection. Reactor siting

criteria. Reporting and recordkeeping

requtiremenls.

Puroiast to the Atomic r.nergy Act of

1954. as amended. the Energy

Reorg:anization Act of 1974. us amended.

and section 552 and 553 of Title 5 (if the

United States Codde. the following

amendments to Title 10. Chapter L C;ode

of Federal Regulations. Ilart 50 are

published as a'document subject to

codification.

PART 50--DOMESTIC UCENSING OF

PRODUCTION AND UTILIZATION

FACILITIES

1. The authority citation for Part 50

continues to read as follows:

Authority: Sees. 1031104.181.182.183,16.

189. 68 StaL 136. 937.948 953.954. 955, M& as

amended. sec. 234.83 Slat 1244. as amended

(42 U.S.C. 2133.2134.221. 2232. 22332236.

2239. 22]2: sees. 201. 202. 206.88 Slat. 1242.

1244.1248& as amended (42 U.S.C. 5841. 5842

58463. unless otherwise noted.

Section S7 also issued under Pub. L 95-

601. sec. 10. 92 Slat. 2951 142 U.S.C. 5851).

Sections 501%.W0.11 and 50.92 also issued

under fhlh. L 97-415.95 Stat. 2073 (42 U.S.C.

22391. Section 5078 also Issued under sec.

122. 68 Slat. 939 142 U.S.C. 2152). Sections .

50.80-5n.81 also Issued under sec. 184. 68 Slat.

954. as amended (42 U.S.C. 2234). Sections

50.100-50.102 also issued under sec. 1i8s G8

Slat. 955 (42 U.S.C. 2238).

For the purposes or sec. 23. 68 Stat. 958. as

amended (42 If.S.C. 22733. 11 50.10 (J). (b).

and (c:. 50.44.50.46, 50.41, 50.54. and 50.A8(al

ar issue.'d under sec. 161b. 68 Slat. 94L. as

amcnded (42 U.S.C. ZZOl(b)l: 150.10 b) and

(c) and 50.54 are Issued under sec. 161L 68

Stat. 949. us amended (42 U.S.C. 22fl1(1): and

I I .55(e). 50W59t1]. 50.70. 50.71. 50.72. and

50.78 are issued under sec. 101o. 68 Stat. 950.

as amended (42 U.S.C. 2201(n)].

2. A new puragaph (z) Is added to

5 50.54 to rend as follows.

150.54 Conditions of licenses.

(z) Fach licensee with a utilization

facility licensed pursuant to sections 103

or 104b. of the Act shall Immediately

miotify the NRC Operations Center of the

occurrence of any event specified In

50.72 S of this part.

3. Section 50.72 is revised to read as

follows:
39046 Federal Register / Vol. 48. No. 168 1 Monday. August 29. 1983 / Rules and Regulations

§ 50.72 Immediate notiflcation

requirements for operating nuclear power

reactors.

(a) General Requirements. I (1),Each

nuclear power reactor under J 50.21(b)

or § 50.22 of this part shall notify the

NRC Operations Center via the

Emergency Notification System of:

(i) The declaration of any of the

Emergency Classes specified in the

licensee's approved Emergency Plan: 'or

(ii) Of those non-Emergency events

specified in paragraph (b) of ths section.

(2) If the Emergency Notification

System is inoperative, the licensee shall

make tlife ]required notifications via

commerical telephone service. other

dedicated telephone system, or any

other method which will ensure that a

report is made as soon as practical to

the NRC Operations Center.*

(3) The licensee shall notify the NRC

immediately after notification of the

appropriate State or local agencies and

not later than one hour after the time the

licensee declares one of the Emergency

Classes.

(4) When making a report under

paragraph (a)(3) of this section. the

licensee shall identify:

(I) The Emergency Class declared- or

(it) Either paragraph (bJ(1). 'One-Hour

Report." or paragraph (b)(2), 'Four-Hour

Report." as the paragraph of this section

requiring notification of the NonEmergency Event.

(b) Non-Emergency Events. (1) OneFHour Reports. If not reported as a

leclaration of an Emergency Class

inder paragraph (a) of this section. the

icensee shall notify the NRC as soon as

3ractical and In all cases within one

iour of the occurrence of any of the

'allowing:

(i)(A) The initiation of any nuclear

ilant shutdown required by the plant's

rechnical Specifications.

(B) Any deviation from the plants

rechnical Specifications authorized

)ursuant to I 50.54(x) of this part.

(i1) Any event or condition during

iperation that results in the condition of

he nuclear powerplant. including Its

,rincipal safety barriers, being seriously

iegraded. orresults in the nuclear

,owerplant being:

(A) In a unanalyzed condition that

ignifIcantly compromises plant safety.

(B) In a condition that Is outside the

eslgn basis of the plant: or

' Other requirements W kimetdi-ia notiefcatom of

is NRC by licensed aperaUq r.dlear power

!actors are contained e•tewlers this dMMp, In

tldgla. I UoLM I 20.4M&I 506. -ad I n=7

'Thes Emergency Clas• I n addresmed In

ppenldi Eea this purt

'Cajunerj lelephome mIr - tho NRC

perstions Center ts (zm] s1i-0om.

(C) In a condition not covered by the

plant's operating and emergency

procedures.

(iii) Any natural phenomenon or other

external condition that poses an actual

threat to the safety of the nuclear

power-plant or significantly hampers

site personnel in the performance of

duties necessary for the safe operation

of the plant.

(iv) Any event that results or should

have resulted in Emergency Core

Cooling System (ECCS) discharge into

the reactor coolant system as a result of

a valid signal

(v) Any event that results in a major

loss of emergency assessment

capability, offsite response capability, or

communications capability (e.g.

significant portion of control room

indicatiom. Emergency Notification

System, or offsite notification system).

(vi) Any event that poses an actual

threat to the safety of the nuclear

powerplant or significantly hampers site

personnel in the performance of duties

necessary for the safe operation of the

nulcear powerplant including fires, toxic

gas releases, or radioactive releases.

(2) Four-Hour Reporis. If not repo'rted

under paragraphs (a) or (b)(1) of this

section. the licensee shall n6tify the

NRC as soon as practical and in all

cases, within four hours of the

occurrence of any of the following:

(I) Any event. found while the reactor

is shutdown. that. had It been found

while the reactor was In operation.

would have resulted in the nuclear

powerplant. including its principal

safety barriers, being seriously degraded

or being in an unanalyzed condition that

significantly compromises plant safety.

(it) Any event or condition that results

in manual or automatic actuation of an

Engineered Safety Feature (ESFI,

including the Reactor Protection System

(RPS). However. actuation of an ESF,

Including the RPS. that results from and

is part of the preplanned sequence

during testing or reactor operation need

not be reported. (iii) Any event or condition that alone

could have prevented the fulfillment of

the safety function of structures or

systems that are needed to:

(A) Shut down the reactor and

maintain it in a safe shutdown

condition.

(B) Remove residual heat.

(C) Control the release of radioactive

materiaL or

(D) Mitigate the consequences of an

accident.

(Iv)(A) Any airborne radioactive

release that exceeds I times the

applicable concentrations of the limits

specified in Appendix B, Table lI of Part

20 of this chapter In unrestricted areu,

when averaged over a time period of

one hour.

(B) Any liquid effluent release that

exceeds 2 times the limiting combined

Maximum Permissible Concentration

(MPC) (see Note I of Appendix B to Part

20 of this chapter) at the point of entry

into the receiving water (i.e..

unrestricted area) for all radionuclides

except tritium and dissolved noble

gases. when averaged over a time period

of one hour. (Immediate notifications

made under this paragraph also satisfy

the requirements of paragraphs (a)(2)

and (b)(2) of 1 20.403 of Part 20 of this

chipter.)

(v) Any event requiring the transport

of a radioactively contaminated person

to an offsIte medical facility for

treatment.

(vi) Any event or situation. related to

the health and safety of the public or

onsite personneL or protection of the

environment. for which a news release

Is planned or notification to other

government agencies has been or will be

made. Such an event may include an

onsite fatality or inadvertent release of

radioactively contaminated materials.

(c) Followup Notification. With

respect to the telephone notifications

made under paragraphs (a) and (b) of

this section. in addition to making the

required initial notification. each

licensee, shall during the course of the

event*

(1) Immediately reporL. (I) any further

degradation in the level of safety of the

plant or other worsening plant

conditions, including those that require

the declaration of any of the Emergency

Classes. if such a declarations has not

been previously made, or (ii) any change

from one Emergency Class to another, or

(ili) a termination of the Emergency

Class.

(2) Immediately reporL (i) the results

of ensuing evaluations or assessments of

plant conditions. (ii) the effectiveness of

response or protective measures taken,

and (fii) Information related to plant

behavior that is not understood.

(3) Maintain an open. continuous

communication channel with the NRC

Operations Center upon request by the

NRC.

Dated: at Washington, D.C. this 23d day of

August. 1982.

For the Nuclear Regulatory Commisslon.

Samuel . Chz'fk.

SecWeOof•the Comnnmisso

BUM CODE 750& Fela4"uh
APPENDIX E

10 CFR 50.73 INCLUDING

STATEMENT OF CONSIDERATIONS

Published in the Federal Register

on July 26, 1983

(Vol. 48, No. 144, pages 33850-33860)

NOTE: This Federal Register notice does not provide a current and correct version of 10

CFR 50.73, which has been amended several times since 1983. Its purpose

here is to present the Statement of Considerations, which explains the basic

reporting requirements of 10 CFR 50.73.

NUREG-1022, Rev. 1

33850 Federal Register / VoL 48. No. 144 / Tuesday. July 28. 1983 / Rules and Regulations

ACnoTN: Final rule.

SUMARr. The Commission is amending

its regulations to require the reporting of

operational experience at nuclear power

plants by establishing the Licensee

Event Report (LER) system. The final

rule Is needed to codify the LER

reporting requirements in order to

establish a single set of requirements

that apply to all operating nuclear

power plants. The final rule applies only

to licensees of commercial nuclear

power plants. The final rule will change

the requirements that define the events

and situations that must be reported.

and will define the information that

must be proyided in each report.

EFECTIVE DAT: January 1 1984. The

incorporation by reference of certain

publicatoions listed In the regulations Is

approved by the Director of the Federal

Register as of January 1. 298

FOR FURTHER INFORMATION COMTACT

Frederick J. Hebdon. Cide f Program

Technology Branch. Office for Analysis

and Evaluation of Operational Data.

US. Nuclear Regulatory Commission.

Washington. D.C. 20555= Telephone (301)

492-4480.

SUPPLEMENTARY INFORMATIOI:

L Background

On May 6.1982. the NRC published in

the Federal Register (47 FR 19543), a

Notice of Proposed Rulemaking that

would modify and codify the existing

Licensee Event Report (LER) system.

Interested persons were invited to

submit written comments to the

Secretary of the Commission by July 6.

2982. Numnerous comments were

received.-After consideration of the

comments and other factors involved.

the Commission has amended the

proposed requirements published for

public comment by clarifying the scope

and content of the requirements.

particularly the criteria that define

which operational events must be

reported.

The majority of the comments on the

proposed rule. (1) Questioned the

meaning and Intent of the criteria that

defined the events which must be

reported. (2) questioned the need for

reporting certain specific types of

events, and (3) questioned the need for

certain information that would be

required to be Included In an LE..

Sedtlon MI of this notice discusses the

comments'ti sf6re detail.

CopWes of the documents an available for public inspection and copying fore fee at the Public

Document Room at M17 H Shvet NW. Washington.

D.C.

EL Rulemaking initiation

The Nuclear Plant Reliability Data

(NPRD) system is a voluntary program

for the reporting of reliability data by

nuclear power plant licensees. On

January 30. 1980 (45 FR 6793).' the NRC

published an Advance Notice of

Proposed Rulemaking that described the

NPRD system and invited public

comment on an NRC plan to make it

mandatory. Forty-four letters were

received in response to the advanced

notice. These comments generally

opposed making the NPRD system

mandatory on the grounds that reporting

of reliability data should not be made a

regulatory r•quiremenL

In December 1980. the Commission

decided that the requirements for

reporting of operational experience data

neededmajor revision and approved the

development of an Integrated

Operational Experience Reporting

(IOER) system. The.IOER system would

have combined. modified. and made

mandatory the existing Uce'csee Event

Report (LER) system and the NPRD

system. SECY 80-507' discusses the

IOER system.

As a result of the Commission's

approval of the concept of an IOER

system. the NRC published another

advance notice on January 15.1981 (46

FR W41). Thds advance notice explained

why the NRC needed operational

experience data and described the

deficiencies in the existing LER and

NPRD systems.

On June 8.1981. the Institute of

Nuclear Power Operations (INPO)

announced that because of Its role as an

active user of NPRDs data it would

assume responsibility for management

and funding of the NPRD system.

Further. INPO decided to develop

criteria that would be used in its

management audits of member utilities

to assess the adequacy of participation

in the NPRD system.

The twvo principal deficiencies that

had previously made the NPRD system

an inadequate source of reliability data

were the inability of its committee

management structure to provide the

necessary technical direction and a low

level of participation by the utilities. The

commitments and actions by INPO

provided a basis for confidence that

these two deficiencies would be

corrected. For example, centralizin8 the

management and funding of NPRDS

withn INPO should overcome the

previous difficulties associated with

management by a committee and

funding from several independent

organizations. Further, with INPO

focusing upon a utility's participatido In

NUCLEAR REGULATORY

COMMISSION

10 CFR Parts 20 and 50

Licensee Event Report System

AGENc,. Nuclear Regulatory

Comnlnission.
Federal Register / VoL 48. No. 144 / Tuesday. July 28. 1983 / Rul,.* and Regulations 338,51

NPRDS as a specific evaluation

parameter during routine management

and plant audit activities, the level of

utility participation, and theiefore, the

quality and quanLtity of 'PRDS data,

should significantly increase. However,

the Commission will continue to have an

active role in NPRDS by participating in

an NPRDS User's Group, by periodically

assessing the quality and quantity of

Information available fom NPRDS, and

by auditing the timely availability of the

lnfozamation to the NRC.

Since there was a likelihood that

NPRDS under INO direction would

meet the NRC's need for reliability data.

it was no longer necessary to proceed

with the IOERS. Hence, the collection of

detailed technicaLdescriptions of

significant events could be addressed in

a separate rulemaking to modify and

.codify the existing LER reporting

requirements. See SECY 81-494 for

additional details concerning IOERS.

However. the Commission wishes to

make It explicitly clear that it is relaxing

the reporting requirements with the

expectation that sufficient utility

participation, cooperation, and support

of the NPRD system will be forthcoming.

If the NPRD system does not become

operational ata satisfactory level in a

reasonable time. remedial action by the

Commission in the forni of additional

rulemaking may beconme necessary.

On October . 1981 the NRC

published an advanced notice (40 FR

49134) that deferred development of the

IOER system and sought public

comment on the scope and content of

the LER system. Six comment letters

were received In response to this

ANPRA. All of the comments received

were reviewed by the staff and were

considered in the development of the

proposed LER rule. See SECY 8-3 ' for

additional details.

This rule Identifies the types of

reactor events and problems that are

believed to be significant and useful to

the NRC in its effort to identify and

resolve threats to public safety. It is

designed to provide the information

necessary for engineering studies of

operational anomalies and trends and

patterns analysis of operational

occurrences. The same information can

also be used for other analytic

procedures that will aid in Identifying

accident precursors.

The Commission believes that the

NRC should continue to seek an

Improved operational data system that

I maximize the value of operational

data. The system should encompass and

Integrate operational data of events and

problem sequences identified in this

rule. NPRDS data. and such other

Information as is required for a

comprehensive integrated analyticallyversatile system.

The Brookhaven Study. published as'

BNL/NUREG 5186M NUREG/C4 3206

discusses data collection and storage

procedures to support multivariate.

multicase analysis. While the range of

reactor configurations In the US.

nuclear industry presents some

methodological and interpretative

problems, these difficulties should not

be insurmountable. The Commission

believes that the NRC should have as a

specific objective the development.

dimonstration, and Implementation of

an integrated system for collecting and

analyzing operational data that will

employ the predictive and analytical

potential of multicase, multivariate

analyses. Accordingly, the staff has

been directed to undertake the work

necessary to develop and demonstrate

such a cost-effective integrated system

of operational data collection and

analyses.

If the design of the system

demonstrates that such a system is

feasible and cost-effective. development

of the system to the point of Initiating

rule should be completed by July 1988

13L Analysis of Comments

The Commission received forty-seven

(47) letters commenting on the proposed

rule. Copies of those letters and a

detailed analysis of the comments are

available for public inspection and

copying for a fee at the NRC Public

Document Room at 1717 H Street. NW.

Washington. D.C. A number of the more

substantive Issues are discussed below.

Licensee Resources

Of particular concern to the

Commission was the impact that the

proposed rule would have on the

resources used by licensees to prepare

LERs. The Commission's goal was to

assure that'he scope of the rule would

not increase the overall level of effort

above that currently required to comply

with the exdsting LER requirements.

Thirty letters of the 47 received

contained comments on the overall

acceptability of the proposed rule or

commented directly on the .question of

scope and/or resources associated with

the proposed rule. The views of the

commenters can be characterized as

follows

1. Five commenters felt that the scope

and level of effort would be greatly

expanded by the proposed rule,

Estimates included an increase of 100

man-year. for the entire industry, an

increase of three times the current effort,

and an increase of $100.000 ad'd 2 manyears annually for each plant

2. Four commenters felt that the leVel

of effort would be increased but not

significantly.

3. One commenter felt that the

proposed rule would have a minimal

effect on the level of effort required.

4. Two commenters felt that the

proposed rule would significantly reduce

the number of LERs filed.

3. Thirteen commenters endorsed the

objective of Improving LER reporting but

felt that changes in the proposed rule

were needed. These commenters did not

directly address the resource issue.

8. Five commenters endorsed the

proposed rule andlor felt that It was a

significant improvement over the

existing reporting requirements.

Based on these comments and Its own

assessment of the impact of this rule, the

Commission has concluded that the

impact of this rule will be no greater

than the impact of the existing LER

requirements, and this rule will not

place an unacceptable burden on the

affected licensees.

Relationship Between the LER Rule

(§ 50.73) and the Immediate Notification

Rule (I 50J72)

As a parallel activity to the

preparation of I 50.73, the Commission

Is amending its regulations (15072)

which require that licensees for nuclear

power plants notify the NRC Operations

Center of significant events that occur at

their plants. On December 21. 19•, the

Commission published in the Federal

Register a proposed rule (48 FR 61894)

that described the planned changes in

§ 502.

The Federal Register notice

accompanying the proposed LER rule

(Le. I 50.73) stated that additional

changes anticipated to 50.72 would be

made but they would be "* *largely

administrative and the revised 50.2

would not be significantly modified nor

would It-be published again for public

comment." Several commenters

disagreed with this conclusion.

The commenters did, however, agree with the Commission's position that

Inconsistencies and overlapping

requirements between the two rules

need to.be eliminated.

The Commission has carefully

reviewed the proposed requirements in

thb LER and Immediate Notification

rules and has concluded that although

changes to both have been made

(largely In response to.public comments)

to clarify the intent of the rules, the

original intent and scope have not been

significantly changed. Therefore, the

Commission has concluded that these

two rule: zeed not be published again

for public comment.
33852 Federal Register / Vol. 481 No. 144 / Tuesday, July 2&, 1983 / Rules and Regulations

Engineering Iudgment

In the Federal Register notice that

accompanied the proposed rule. the

Commission stated that licensee's

engineering Judgment may be used to

decide if an event Is repoitable. Several

commenters expressed the belief that

some wording should be added to the

rule of reflect that the NRC will also rt'e

judgment in enforcement of this

regulation where the licensee Is

requested to use engineerg Judgment.

The Commission believes that the LER

rule adequately discusses the need for

and application of the concept of

"engineering judgment" The concept

itself includes the recognition of the

existence'of a reasonable range of

Interpretation regarding this rule. and

consequently the Commission

recognizes and hereby acknowledges

the need for flexibility in enforcement

actions associated with this rule. The

Commission believes that this concept Is

sufficiently dear and that additional

explicit guidance Is not necessary.

Rgeporting Schedule

In the Federal Regster notice that

accompanied the proposed rule. the.

Commission stated that it had notyet

decided If the reports should be

submitted in fifteen days or thirty days

following discovery of a reportable

event. Many commenters stated that the

time frame for reporting LERs should not

be less than thirty days after the

discovery of a reportable event.

One commenter estimated the impact

of a requirement to submit a report

sooner than 30 days following discovery

of a reportable event would be an

increase of approximately 40 man years

per year for the currently operating

plants. In addition the commenter

estimated that If a summary report were

also required the reporting burden

would increase an additional 12 man

years for the currently operating plants.

In response to these comments, the

Commission has decided to require that

LERs be submitted within 30 days of

discovery of a reportable event or

situation.

Reporting of Reactor TLips

Section 50.73(a)(1) of the proposed

rule (I 50.73(a)(2)(iv) of the final rule)

required reporting of any event which

results in an unplanned manual or

automatic actuation of any Engineered

Safety Feature (ES?) including the

Reactor Protection System (RPS). Many

commenters agreed that these events

should be trended and analyzed. but

disagreed that they deserve to be

singled out as events of special

sitiificance (I e.. dvents reportable as

LEM). They noted that reports of RPS

actuations are already reported to the

NRC In the Monthly Operating Status

Report, as well as telephoned to the

NRC Operations Center.

In addition, the Institute of Nuclear

Power Operations (INPO) analyzed the

frequency of reactor scrams during a

one-month period. This analysis

Indicated that an average of 55 reactor

trips would be reportable each month

under the proposed rule. INPO equated

this to 680 additional LERs per year for

all currently operating plants, or

approximately 32 man-years of

additional effort for all the currently

operating plants based ulion the

assumption that each LER requires 100

man-hours of effort to prepare and

analyze.

The Commission still believes/hat

ESF actuations. includ.i." reactor trips.

frequenty are associated with

significant plant transients and are

indicative of eventi that are of safety

significance. In additio6, if the ESFs are

being challenged during routine

transients, that fact is of safety

significance and should be reported.

In addition, the Commission does not

agree with the estimate that each LER

submitted for a routine reactor trip

would require. on the average. 100 manhours to prepare and analyze. Licensees

are already required to make internal

evaluation of and document significant

events. Including reactor trips.

Therefore. the Incremental impact of

preparing and analyzing the LER should

be significantly less than 100-man hours.

In addition, the actual increase in

burden would be offset by reductions In

the burden of reporting less significant

events that would no longer be

reportable.

Coordinaton With Other Reporting

Requhi"uments

Several commenters noted. that the

proposed rule did not appear to be

coordinated with other existing

reporting requirements, and that

duplication of licensee effort might

result. They recommended that LER

reporting be consolidated to eliminate

potential duplication of other existing

reporting requirements.

The Commission has reviewed

existing NRC reporting requirements

(e.g., 10 CFR Parts 20 and 21. 1 50.U5(e),

1 52.71 .SM. 3 731.M and NUREG0654) qndlias attempted, to the extent

practicable, to eliminate redundant

reporting and to ensure that the-various

repodingreqfalrements are consistent.

Ilany of the changes In the final LER

rile are as a rer,!, of this effort. These

changes resulted In extensive revisions

in the wording of criteria contained in

this rule. but did not change the original

scope of Intent of the requirements. In

addition. in order to make the

requirements in 1150.72 and 50.73 more

compatible. the order (Le.. numbering) of

the criteria in 150.73 has been changed.

The changes are noted In the discussion

of each paragraph below.

Finally, conforming amendments are

being made to various sections of Parts

20 and 50 in order to reduce the

redundancy in reporting requirements

that apply to operating nuclear power

plants. In general, these amendments

wll reqle that:

1. Licensees that have an Emergency

Notification System (ENS) make the

reports required by the subject sections

via the ENS. All other licensees will

continue to maue the reports to the

Administrator of the appropriate NRC

Regional Office.

2. Written reports required by the

subject sections be submItted to the

NRCDocument Con*trol Desk In

Washington. D.C.. with a copy to the

appropriate Regional Offices.

3. Holders of licenses to operate a

nuclear-power plant submit the written

reports required by the subject sections

in accordance with the procedures

described in I 50.3(b).

The criteria contained In the subject

sections which define a reportable event

have not been modified.

Similar changes are also planned as

part of curent activities to make more

substantive changes to Part 22L

I 50.55(e). and I 73.71.

Nonconservative Interdependence

Several commenters expressed

difficulty In understanding the meaning

of the phrase "nonconservative

interdependence" as used in the

proposed I 50.73(a)(3). The wording of

I 50.73(a)(3) (1 50.73(a)(2)(vii) of this

final rule) has been changed to eliminate

the phrase "non conservative

interdependence" by specifically

defining the types of events that should

be reported. The revised paragraph does

not, however, change the intent of the

original paragraph.

Sabotage and Threats of Violence

Several commenters noted that the

security-related reporting requirements

of 1 50n7(a)(6) (1 50(a)(2)(ill) of this

final rule)) were already contained in

greater detail In 10 CFR 73. For

instance. 173.71 requires an act of

sabotage to be ieported Immedlatdly,

followed by a written report within 15

days. The proposed rule would have

required an LER to be filed within 30

days. Although distribution ofreports is

somewhat different redundant reporting
Federal Register / VoL 48. No. 144 / Tuesday. July 28. 1983 / Rules and Regulations 33M

would have occurred. The commenters

recommended that the Commission

ensure consistency between 1150.73 and 73.71.

In response to these comments the

Commission has deleted the reporting of

sabotage and threats of violence from

1 50.73 because these situations are

adequately covered by the reporting requirements contained in I 73.71.

Evacuation of Rooms or Buildings

Many commenters stated that the

reporting of In-plant releases of

radioactivity that require evacuation of

Individual rooms (I 5023(a)(7) In the

proposed rule or (I 50.3(a2)(x) of this

final rule) was Inconsistent with the

general thrust of the rule to require

reporting of significant events. They noted that minor spills. small gaseous

waste releases, or the disturbance of

contaminated particulate matter (e.g.,

dust) may all require the temporary

evacuation of Individual rooms until the

airborne concentrations decrease or

until respiratory protection devices are

utilized. They noted that these events

are fairly common and should not be

reportable unless the required

evacuation affects the entire facility or a

major portion thereof.

In response to these comments the

wording of this criterion (1 50.aXZ)(x)

In the final rule) has been changed to

significantly narrow the scope of the

criterion to Include only those events

which significantly hamper the ability of

site personnel to perform safety-related

activities (e.g4 evacuation of the main

control room).

Energy Industry Identicatton System

Many commenters noted that the

requirement to report the Energy

Industry Identification System (BUS)

component function Identifier and

system name of each component or

system referred to In the LER

description would be a significant

burden on the licensee.

They suggested instead that the

NPRDS component Identiflers be used in

place of the MIS component identifiers

which are not yet widely used by the

Industry.

The Commission continues to believe

that EIS system names and component

function identifiers are needed In order

that LERm from different plants can be

compared. We do not. however, suggest

that the EIIS Identifiers be used

thriughout the plant, but only that they

be added to the LER as it Is written. A

simple, Inexpensive table cotdd be used - to translate plant Identifiers into

equivalent EIIS identifiers.

The Commission considered the

system and component ldanti!rs used

In NPRDS as an alternative. It Is our

understanding, however, the NPRDS will

soon adopt the ES system titles. so a

distinction should no longer exist In

addition. LERs frequently include

systems that are not included ift the

scope of NPRIJS (Le., an NPRDS systim

identification does not exist) while ElIS.

on the other hand. includes all of the

systems commonly found: in commercial

nuclear power plants. Further. NPRDS

includes only 39 component Identifiers

(e.g, valve, pump). The Commission

believes that this limited number does

not provide a sufficiently detailed.

description of the component function

involved.

Function of Failed Components and

Status of Redundant Components

Many commenters said that

information required In (I 50.73(b)(2) (vi)

and (vii) of the proposed rule should not

be a requirement In the LER. They

argued that this Information is readily

available In documents previously

submitted to the NRC by licensees and

are available for reference.

The final rule (I 073(b)(2)(IXG)) has

been modified to narrow the scope of

the Information requested by the

Commission.

While this general. information may be

available in licensee documents

previously submitted to the NRC. the

Commission believes that a general

understanding of the event and its

significance should be possible without

refeience to additional. documentation

which may not be readily or widely

available, particularly to the public.

The Commission continues to believe

that the licensee should prepare an LER

in suffclentdepth so thiat

knowledgeable readers who are

conversant-with the design of

commercial nuclear power plants, but

are not familiar with the details of a

particularplant, can understand the

general characteristics of the event (e.g..

the cause, the Ificance, thecorrective action). As. suggested bythe

commenters, more detailed Information

to support engineerlng evaluations and

case studies will be obtained, as

needed, directly from the previously

submitted licensee documents.

Engneering Eravakair,

The overview discussion of the

proposed rule contains, the following

statement: "If the NRC staff decides that

the event was especially significant

from the standpoint of safely, the staff

may request that the licensee perform

an engineering evaluation of the event

and.describe the results of that

evaluation."

Several qommenters argued that the

inclusion of the requirement that the

licensee perform an engineering

evaluation of certain events at the staffs

request appeared unjustified and would

add substantially to the burden of

reporting. They argued that the licensee

should be required to submit only the

specific additional information required

for the necessary engineering evaluation

rather than to perform the evaluation.

The rule has been modified to require

only the submittal of any necessary

additional Information requested by the

Commission in writing.

IV. Specific Findings

Overvdew of the LER System

When this final LER rule becomes

effective, the LER will be a detailed

narrative description of potentially

significant safety events. By describing

in detail the event and the planned

corrective action, It will provide the

basis for the careful study of events or

conditions that might-lead to serious

accidents. If the NRC staff-decides that

the event was especially significant

from the standpoint of safety, the staff

may request that the licensee provide

additional information and data

associated with the event.

The licensee will prepare an LER for

those events or conditions that meet one

or more of the criteria contained In

I 50.73(a). The criteria are based

primarily on the nature. course, and

consequences of the event.Therefore.

the final LER rule requires that events

which meet the criteria are to be

reported regardless of the plant

operating mode or power leveL and

regardless of the safety significance of

the components, systems, or structures

involved. In trying to develop criteria for

the Identification of events reportable as

LEts. the Comjmission has concentrated

on the potential consequences of the

eventas the measure of significance.

Therefore. the reporting criteria, in

generul, do not specifically address

classes of Initiating events or causes of

the event. For example, there IsMd

requirement that all personnel errors be

reported. However, many reportable

events will involve or have been

initiated by personnel errors

Finallyjit should be noted that

licensees are permitted and encouraged

to report any event that does not meet

the criteria contalned in 1 50.3a). if the

licensee believes that the event might be

of safety significance, or of generic

interest or concerReporting

requirements aside. assurance of safe

operation of all plants depends on

accurate and complete reporting by each
333H Federal Register / VoL 48. No. 144 / Tuesday. July 2e, 1983 1 Rules and Regulations

licensee of all events having potential

safety significance.

Paragraph-by-Pargraiph Explanation of

the LER Rule

The significant provisions of the final

LER rule are explained below, The

explanation follows the order In the

proposed rule.

Paragraph S0.73(aJ(2)(iv) (proposed

paragraph 50.73(a)(1)) requires reporting

of: "Any event or condition that resulted

In manual or automatic actuation of any

Engineered Safety Feature (SFj

including the Reactor Protection System

(RPS). However, actuation of an ESF.

including the RPS. that resulted from

and was part of the preplanned

sequence durifg testing or reactor

operation need not be reported.

This paragraph requires-events to be

reported whenever an ESP actuates

either manually or automatcally,

regardless of plant status. It Is based on

the premise that the ESFs are provided

to mitigate the consequenc"s of a

significant event and, therefore: (1) They

should work propetly when called upon.

and (2) they should not be challenged

frequently or unnecessarily. The

Commission is interested both in events

where an ESF was needed to mitigate

the consequences (whether or not the

equipment performed properly) and

events where an ESF operated

unnecessarily.

"Actuation" of multichannel ESF

Actuation Systems is defined as

actuation of enough channels to

complete the minimum actuation logic

(I.e, activation of sufficient channels to

cause activation of the ESF Actuation

System). Therefore, single channel

actuations& whither caused by failures

or otherwise, are not reportable if they

do not complete the minimum actuation

logic. Operation of an ESF as part of a

planned operational procedure or test

(e.g.. startup testing) need not be

reported. However, if during the planned

operating procedure or test, the ESF

actuates in a way that is not part of the

planned procedure, that actuation must

be reported. For example, if the normal

reactor shutdown procedure requires

that the control rods be Inserted by a

manual reactor trip, the reactor trip need

not be reported. However, If conditions

develop during the shutdown that

require an automatic reactor trip, such a

reactor trip must be reported.

The fact that die safety analysis

assimes that an ESF will actuate

automatically during certain plant

conditions does not eliminate the need

to report that actuation. Actuations that

need not be reported are those Initiated

for rmasons other than to mitigate the

consequences of an event (e.g., at the

discretion of the licensee ai part of a

planned procedure or evolution).

Sections 50.73(a)(2) (v) and NvO)

(proposed I 50.73(a)(2)) require reporting

of:

(v} Any event or condition that alone could

have prevented the fulfillment of the safety

function of structures or systems that are

needed to:

(A) Shut down the reactor and maintain It

in a safe shutdown condition:

(B) Remove residual heat;

(C) Control the release if radioactive

materl: or

(D) Mitiate the consequences of an

accident..

(vi) Events covered in paragraph (aXZXv)

of this section may Include one or more

personnel errors equipment failures, and/or

discovery of design. analysis, fabrication.

construction. andlor procedural

Inadequies. However. Individual

om t failures neednot be reported

pursuant to this paragraph If redundant

equipment In the same system was operable

and available to perform the required safety

finction.

The wording of this paragraph has

been changed from the proposed rule to

make it easier to read. The Intent and

scope of the paragraph have not been

changed.

The Intent of this paragraph Is to

capture those events where there would

have been a failure of a safety system to

properly complete a safety function.

regardless of when the failures were

discbvered or whether the system was

needed at the time.

This paragraph is also based on the

assumption that safety-related systems

-and structures are intended to mitigate

the consequences of an accident. While

I 50.73(a)(2)(lv) of this final rule applies

to actual actuations of an ESF.

I 50.73(a)(2)(v) of this final rule covers

an event or condition where redundant

structures, components, or trains of a

safety system could have failed to

perform their Intended function because

of: one or more personnel errors.

Including procedure violations;

equipment failures; or design. analysis,

fabrication. construction, or procedural

deficiencies. The event must be reported

regardless of the situation or condition

that caused the structure or systems to

be unavailable, and regardless of

whether or not an alternate safety

system could have been used to perform

the safety function (eg.. High Pressure

Core Cooling failed, but feed-and-bleed

orLow Pressire Core Cooling were

available't6 pf6vide the safety function

of.jore coolin).

The applicability of this paragraph

includes those safety systems designed

to mitigate the consequences of an

accident (e.g.. containment Isolation.

emergency filtration). Hence, minor

operational events involving a specific

component such as valve packing leaks,

which could be considerd a lack of

control of radioactive material, should

not be reported under this paragraph.

System leaks or other similar events

may. however, be reportable under other

paragraphs.

It should be noted that there are a

limited number of single-train systems

that perform safety functions (eg., the

High Pressure Coolant Injection System

in BWRe). For such systems, loss of the

single train would prevent the

fulfillment of the safety function of that

system and. therefore, must be reported

even though the plant Technical

Specifications may allow such a

condition to exist for a specified limited

length of time.

It should also be noted that. If a

potentially serious human error Is made

that could have prevented fulfillment of

a safety function. but recovery factors

resulted In the error being corrected, tne

error is still reportable.

The Commission recognizes that the

application of this and other paragraphs

of this section involves the use of

engineering judgment on the part of

licensees. In this case. a technical

judgment must be made whether a

failure or operator action that did

actually disable one train of a safety

system, could have. but did not,. affect a

redundant train within the ESF system.

If so, this would constitute an event that

"could have prevented" the fulfillment

of a safety function. and, accordingly,

must be reported.

If a component falls by an apparently

random mechanism it may or may not

be reportable if the functionally

redundant component could fadl by the

same mechanism. Reporting Is required

If the failure constitutes a condition

where there Is reaionable doubt that the

functionally redundant train or channel

would remain operational until It

completed its safety fuinction or is

repaired. For example, if a pump In one

train of an ESF system falls because of

improper lubrication. and engineering

judgment indicates that there is a

reasonable expectation that thefunctionally redundant pump in the

other train, whichlwas also improperly

lubricated, would have also failed

before it completed its safety function.

then the actual failure Is reportable and

the potential failure of the functionally

redundant pump must be discussed in

the LER.

For safety systems that include three

or more trains, the failure of two or more

trains should be reported M in the
Federal etister / Vol. 48, No. 144 / Tuesday, July 28, 1983 / Ruwes and Regulations 33855

judgement of the licensee, the functional

capability of the overall system was

jeopardized.

Interaction between systems.

particularly a safety system and a nonsafety system. is also included in this

criterion. For example, theCommission

is increasingly concerned about the

effect of a loss or degradation of what

had been assumed to be non-essential

inputs to safety systems. Therefore, this

paragraph also includes those cases

where a service (e.g.. heating.

ventilation, and cooling) or input (e.g..

compressed air) which is necessary for

reliable or long-term operation of a

safety system is lost or degraded. Such

loss or degradation is reportable if the

proper fulfillment of the safety function

is not cannot be assured. Failures that

affect inputs or services to systems that

have no safety function need not be

reported.

Finally the Commission recognizes

that the licensee may also use

engineering judgment to decide when

personnel actions could have prevented

fulfillment of a safety function. For

example, when an individual improperly

operates or maintains a component, he

might conceivably havemade the same

error for all of th" functionally

redundant components (e.., If he

incorrectly calibrates one bistable

amplifier in the Reactor Protection

System. he could conceivably incorrectly calibrate all bistable

amplifiers). However, for an event to be

reportable It Is necessary that the

actions actually affect or involve

components in more than one train or

channel of a safety system, and the

result of the actions must be undesirable

from the perspective of protecting the

health and safety of the public. The

components can be functionally

redundant (e.g. two pumps in different

trains) or not functionally redundant

(kg., the operator correctly stops a pump

in Train "A" and, instead of shutting the

pump discharge valve in Train "A." he

mistakenly shuts the. pump discharge

valve in Train "W1.

Section 50.73(a)(2)(vii) (proposed I 50.73(aX31} require the reporting ofi

"Any event where a single cause or

condition caused at least one

independent train d? channel to become

inoperable in multiple systems or two

independent trains channels or to

become inoperable in a system designed

to:

(A) Shut down the reactor and

maintain it in a safe shutdown

condition.

(BM Remove residual heat,

(C) Control the release of radioactive

materlah or

(13 Mitigate the consequences of an

accident."

This paragraph has been changed t3

clarify the intent of the ph., ase, "nonconservative interdependence;'?

Numerous comment letters expressed

difficulty in understanding what-this

phrase meant: so the paragraph has

been changed to be more specific. The

new paragraph is narrower in scope

than the original paragraph because the

term Is specifically defined, but the

basic intent is the same.

This paragraph requires those events

to be reported where asingle cause

produced a component or group of

components to become inoperable In

redundant or independent portions (Lae.

trains or channels) of one or more

systems having a safety function. These

events can Identify previously

unrecognized common cause failures

and systems interactions. Such failures

can be simultaneous failures which

occur because ofa single initiating

cause (La., the single cause or

mechanism serves as a common input to

the failures); or the failures can be

sequential (I.e., cascade failures), such

as the case where a single component

failure results in the failure of one or

more additional components.

To be repoxtable" however, the event

or failure mst result in or involve the

failure of independent portions of more

than one train or channel In the same or

different systems. For example, If a

cause or condition caused components

in Train "AX and "1" of a single system

to become inoperable. even if additional

trains (e.g., Train."C") were still

available. the event must be reported. In

addition, if the cause or condition

caused components in Train "A" of one

system and In Train 'I/ of another

system (Ua., a train that Is assumed In

the safety analysis tor be independent) to

become inoperable, the event must be

reported. However, If a cause or

condition caused components in Train

"A" of one system and Train "A' of

another system (La;. trains that are not

assumed In the safety analysis to be

independent), the event need not be

reported unless it meets one or more of

the other criteria in thii section.

In addition, this paragraph does not

include those cases where one train of a

system or a component was removed

from service as part of a planned

evolutioi, In accordance with an

approved procedure, and In accordance

with the plant's Technical

Specifications. For example. if the

licensee removes part of a system from

service to perform maintenance. and the

Technical Specifications permit the

resulting configuration, and the system

or. component Is returned to service

within the time limit specified In the

Technical Specifications. the action

need not be reported under this

paragraph. However. if. while the train

or component 1s out of service, the

licensee identifies a condition that could

have prevented the whole system from

performing its intended function (e.g..

the licensee finds a set of relays that is

wired incorrectly), that condition must

be reported.

Section 5C.3(a)(2)(iJ (proposed

I 50.73(aX(4)) requires reporting ofi

"(A) The completion. of any nuclear

plant shutdown required by the plants

Technical Specifications: or

"(B) Any operation prohibited by the

planes Technical Specifications; or

"(C) Any deviation from the plantes Techcal Specifications authorized

pursuant to I 5 4(x) of this part."

This paragraph has been reworded to

more clearly define the events that must

be reported. In addition, the scope has

been changed to require the reporting of

events or conditions "prohibited by the

planes Technical Specifications" rather

than events where "a plant Technical

Specification Action Statement is not

met." This change accommodates plants

that do not have requirements that are

specifically defined as Action

Statements.

This paragraph now requires events to

be reported where the licensee is

required to shut down the plant because

the requirements of the-Technical

Specifications were not met. For the

purpose. of this paragraph.. shutdown"

is defined as the point in time where the

Technical Specifications require that the

plant be In the first shutdown condition

required by a imiting Condition for

Operation (e.$. hot standby 0Mo& 3) for

PWRs with the Standard Technical

Specifications). If the condition is

corrected before the time limitfor being

shut down (La., before completion of the

shutdown), the event need not be

reported.

In addition. if a condition that was

prohibited by the Technical

Specifications existed for a perod of

time longer than that pernittedby the

Technical Specifications. It must be

reported even If the condition was not

discovered until after the allowable tme

had elapsed and the condition was

rectified immediately after discovery.

Section 5073(a)(2J(ll) (proposed

I 5073(a)(5]) requires reporting of: "Any

event or condition that resulted In the

condition of the nuclearpower plant,

incliding its principal safety barriers,

being seriously degraded. or that

resulted in the nuclearpowerplant

being:
33656 Federal Register I Vol. 48, No. 144 / Tuesday, July 28. 1983 / Rules and Regulations

"(A) In an unanalyzed condition that

significantly compromised plant safety;,

"(B) In a condition that was outside

the design basis of the plant; or

"(C) In a condition not covered by the

plant's operating and emergency

procedures."

This paragraph requires events to be

reported where the plant, including its

principal safety barriers, was seriously

degraded or in an unanalyzed condition.

For example, small voids in systems

designed to remove heat from the

reactor core which have been previously

shown through analysis not to be safety

significant need not be reported.

However, the accumulation of voids that

could Inhibit the ability to adequately

remove heat from the reactor core,

particularly under natural circulation

coriditions, would constitute an

unanalyzed condition and must be

reported. In addition, voiding In

Instrument lines that results In an

erroneous indication causing the

operator to significantly misunderstand

the true condition of the plant is also.an

unanalyzed condition and must be

reported.

The Commission recognizes that the

licensee may use engineering judgment

and experience to determine whether an

unanalyzed condition existed. It Is not

intended that this paragraph apply to

minor variations in individual

parameters, or to problems concerning

single pieces of equipment. For example.

at any time. one or more safety-related

components may'be out of service due

to testing, maintenance, or a fault that

has not yet been repaired. Any trivial

single failure or minor error in

performing surveillance tests could

produce a situation in which two or

more often unrelated, safety-related

components are out-of-service.

Technically, this is an unanalyzed

condition. However, these events should

be reported only if they involve

functionally related components or if

they significantly compromise plant

safety.

Finally, this paragraph also includes

material (e.g. metallurgicaL chemical)

problems that cause abnormal

degradation of the principal safety

ha-riers (Le. the fuel cladding, reactor

coolant system pressure boundary, or

the containment).

Additional examples of situations

included in this paragraph are:

(a) Fuel cladding failures in the

reactor or in the storage pool that

exceed expected values, that are unique

or widespread, or that resulted from

unexpected factors.

(b) Reactor coolant radioactivity

levels that exceeded Technical

Spetification limits for Iodine spikes or,

radioactivity levels at a BWR air ejector

monitor that exceeded the Technical

Specification limits.

(c) Cracks and breaks in piping, the

reactor vessel. or major components in

the primary coolant circuit that have

safety relevance (steam generators.

reactor coolant pumps, valves. etc.).

(d) Significant welding or material

defects in the primary coolant system.

(e) Serious temperature or pressure

transietits (e.g.. transients that violate

the plant's Technical Specifications).

(0 Loss of relief and/or safety valve

operability during test or operation

(such that the number of operable

valves or man-way closures is less than

required by the Technical

Specifications).

(g) Loss of containment function or

integrity (e:g., containment leakage rates

exceeding the authorizedrlimlts).

Section 60.72(a)(2)(ill) (proposed

I 503(a)()) requires reporting of: "Any

natural phenomenon or other external

condition that posed aft actual threat to

the safety of the nuclear power plant or

significantly hampered site personnel in

the performance of duties necessary for

the safe operation of the nuclear power

plant."

This paragraph has been reworded to

make it clear that it applies only to acts

of nature (e.g., tornadoes) and external

hazards (e.g., railroad tank car

explosion). References to acts of

sabotage have been removed because

they are covered by § 73.1. In addition.

threats-to personnel from internal

hazards (e.g. radioactivity releases) are

now covered by a separate paragraph

(I 50.73(a)(2)(x)).

This paragraph requires those events

to be reported where there is an actual

threat to the plant from an external

condition or natural phenomenon. and

where the threat or damage challenges

the ability of the plant to continue to

operate in a safe manner (including the

orderly shutdown and maintenance of

shutdown conditions).

The licensee is to decide If a

phenomenon or condition actually

threatened the plant. For example, a

minor brush fire in a remote area of the

site that was quickly controlled by fire

fighting personnel and. as a result, did

not present a threat to the plant need

not be reported. However, a major forest

fire. large-scale flood, or major

earthquake that presents a clear threat

to the plant must be reported. Industrial

or transportation accidents that

oc-urred near the'ife and created a

plant saf6ty doncern must also be

reported.

Section 50.73(a)(2)(x) (proposed

S (a)(7)) requires reporting oD "Any

event that posed an actual threat to the

safety of the nuclear power plant or

significantly hampered site personnel in

the performance of duties necessary for

the safe operation of the nuclear power

plant including fires, toxic gas releases,

or radioactive releases."

This paragraph has been reworded to

include physical hazards (internal to the

plant) to personnel (e.g., electrical fires).

In addition. in response to numerous

comments, the scope has been narrowed

so that the hazard must hamper the

ability of site personnel to perform

safety-related activities affecting plant

safety.

In-plant releases must be reported if

they require evacuation of rooms or

buildings containing systems Important

to safety and. as a result. the ability of

the operators to perform necessary

safety functions is significantly

hampered. Precautionary evacuations of

rooms and buildings that subsequent

evaluation determines were not required

need not be reported.

Proposed I 50.M(a)(8) was intended to

capture an event that Involved a

controlled release of a significant

am6unt of radioactive material to off•lte

areas. In addition. "significant" was

based on the plant's Technical

Specification limits for the release of

radioactive material. However. this

section has been deleted because the

reporting of these events Is already

required by I 50,73(a)(2)(0) and I 20.405.

Section 50."(a)(2) (viii) and (ix)

(proposed I 50.73(a)(9)) require reporting

of:

(vUiXA) Any airborne radioactivity release

that exceeded 2 times the applicable

concentrations of the limits specified In Table

U of Appendix B to Part W0 of this chapter in

unrestricted areas. when averaged over a

time period of one hour.

(B) Any liquid effluent release that

exceeded 2 times the limiting combined

Maximum Permissible Concentration (MPC)

(see Note I of Appendix B to Part 20 of this

chapter) at the point of entry into the

receiving water (Ie. unrestricted arealfor all

radionuclides except tritium and dissolved

noble gases. when averaged overa time

period of one hour.

(Ix) Reports submitted to the Commission

in accordance with paragraph (aX2)(vfll) of

this section also meet the effluent release

reporting requirements of paragraph

20.405(s)(51 of Part 20 of this chapter.

Paragraph (viii) has been changed to

clarify the requirements to report

releases of radioactive material. The

paragraph is similar to 120.405 but

places a lower threshold for reporting

events at commercial power reactors.

The lower threshold Is based on the

significance of the breakdown of the
Federal Register / VoL 48. No. 144 / Tuesday. July 26, 1983 / Ruiis and Regulations 338S7

licensee's program necessary to have a

release of this size, rather than on the

significance of the impact of the actual

release.

Reports of events covered by

I 50.73(a)(2)(viii) are to be made in lieu

of reporting noble gas releases that

exceed 10 times the instantaneous

release rate. without averaging over a

time period, as implied by the

requirement of I 20.405(a)(5).

Paragraph 50.73(b) describes the

format and content of the LER. It

requires that the licensee prepare the

LER In sufficient depth so that

knowledgeable readers conversant with

thd design of commercial nuclear power

plants. but not familiar with the details

of a particular plant. can understand the

complete event (La., the cause of the

event. the plant status before the event.

and the sequence of occurrences during

the eyent).

Paragraph 50.73(bXl) requires that the

licensee provide a brief abstract

describing the major occurrences during

the event. including'all actual

component or system failures that

contributed to the event. all relevant

operator errors or violations of

procedures, and any significant

corrective action taken or planned as a

result of the event. This paragraph is

needed to give LER data base users a

brief description of the event in order to

identify events of interest.

Paragraph 50.73(b)(2) requires that the

licensee include in the LER a clear.

specific narrative statement of exactly

what happened during the entire event

so that readers not familiar with the

details of a particular plant can

understand the event The licensee

should emphasize how systems,

components, and operating personnel

performed. Specific hardware problems

should not be covered in excessive

detail. Characteristics of a plant that are

unique and that influenced the event

(favorably or unfavorably) musi be

described. The narrative must also

describe the event from the perspective

of the operator (e.g., what the operator

saw. did, perceived, understood, or

misunderstood).

Paragraph 50.73(bj)(3) requires that the

LER Include i summary assessment of

the actual and potential safety

consequences and implications of the

event. This assessment may be based on

the conditions existing at the time of the

event. The evaluation must be carried

out to the extent necessary to fully

assess the safety consequences and

safety margins associated with the

event. An assessment of the event under

alternative conditions must be included

If the incident would have been more

severe (e.g., the plant would have been

In a condition not analyzed in the Safety

Analysis Report) under reasonable and

credible alternative conditions, such as

power level or operating ruode. For

example.'d an event occurred' •hile tije

plant was at 15% power and the same

event could have occurred while the

plant was at 100% power, and. as a

result the consequences would have

been considerably more serious, the

licensee must assess and report those

consequences.

Paragraph 50,73(b)(4) requires that the

licensee describe in the LER any

corrective actions planned as a result of

the event that are known at the time the

LER Is submitted, including actions to

reduce the probability of similar events

occu•ring in the future. After the initial

LER Is submitted only substantial

changes in the corrective action need be

reported as a supplemental LER.

Paragraph 5.37(c) authorizes the NRC

staff to require the licensee to submit

specific supplemental Information

beyond that required by J 50.73(b). Such

information may be required if the staff

finds that supplemental material ts

necessary for complete understanding of

an unusually complex or significant

event. Such requests for supplemental

information must be made in writing.

and the licensee must submit the

requested information as a supplement

to the Initial LER within a time period

mutually agreed upon by the NRC staff

and the licensee.

Paragraph 50.73(1) gives the NRCs

Executive Director for Operations the

authority to grant case-by-case

exemptions to the reporting

requirements contained in the LER

system. This exemption could be used to

limit the collection of certain data in

those cases where full participation.

would be unduly difficult because of a

plant's unique design or circumstances.

Paragraph 50.73(g) states that the

reporting requirementscontained in

J 50.73 replace the reporting

requirements In all nuclear power plant

Technical Specifications that are

typically associated with Reportable

Occurrences.

The reporting requirements

superseded by 150.73 are those

contained in the Technical+Specification

sections that are usually titled "Prompt

Notification with Written Followup"

(Section 0.9.1.8) and "Thirty Day Written

Reports" (Section 6.9.1.9). The reporting

requirements that have been superseded

are also described In Regulatory Guide

1.1e, Revision 4. "Reporting of Operating

Information-Appendix A Technical

Specification." Paragraph 2. "Reportable

Occurrences." The special report

typically described in Section 6.9.2

"Special Reports" of the Technical

Specifications are still required.

V. Regulatory Analysis

The Commission has prepared a

regulatory analysis for this final rule.

The analysis examines the costs and

benefits of the alternatives considered

by the Commission. A copy of the

regulatory analysis Is available for

inspection and copying for a fee at the

NRC Public Document Room. 1717 H

Street. N.W., Washington. D.C. Single

copies of the analysis may be obtained

from Frederick J. Hebon. Chiet Program

Technology Branch. Office for Analysis

and Evaluation of Operational Data.

U.S. Nuclear Regulatory Commission.

Washington. D.C. 20555 Telephone (301)

492-4480.

VL Paperwork Reduction Act Statement

The Nuclear Regulatory Commission

has submitted this rule to the Office of

Management and Budget for such

review as may be appropriate under the

Paperwork Reduction Act. Pub. L 90-

511. The date on which the reporting

requirements of this rule become

effective reflects inclusion of the 60-day

period which the Act allows for such

review.

VII. Regulatory Flexibility Certification

In abcordance with the Regulatory

Flexibility Act of 1980,5 U.S.C. 605(b),

the Commission hereby certifies that

this rule will n6t have a significant

economic impact on a substantial

number of small entities. This final rule

affects electric utilities that are

dominant in their respective service

areas and that own and operate nuclear

utilization facilities licensed under

sections 103 and 104b of the Atomic

Energy Act of 1954. as amended. The

amendments clarify and modify

presently existing notification

requirements.

Accordingly, there is no new,

significant economic Impact on these

licensees, nor do these licensees fail

within the scope of the definition of

"small entities" set forth In the

Regulatory Flexibility Act or the Small

Business Size Standards set out in

regulations issued byithe Small Business

Administration at 13 CFR Part 121.

Ust of Subjects

10 CFR Par20

Ucensed material. Nuclear power

plants and reactors, Penalty. Reporting

and recordkeeping requirements.

20 CFR PARTS 50

Incorporation by reference, Antitrust.

Classified Information. Fire protection.
338SB-- Federal Register / Vol. 48& No. 144 / Tuesday, July =8, 1983 / Rules and Regulations

Intergovernmental relations, Nuclear

power plants and reactors. Penalty.

Radiation protedtion. Reporting and

recordkeeping requirements.-

Under the authority of the Atomic

Energy Act of 1954. as amended, the

Energy Reorganization Act of 1974, as

amended, and 5 U.S.C. 552 and 553, the

rollowing amendments to 10 CFR Parts

20 and 50 are published as a document

subject to codification.

PART 50--DOMESTIC UCENSING OF

PRODUCTION AND UTILIZATION

FACILITIES

1. The authority citation for Part 50

continues to read as follows.

Authorit. Secs. 103,104. 1. 182.1 = 3.1 .

189. 6 Stat. 93. 937. M48. 953.954. 95G=. 95. as

amended. sec. 234. 83 Stat. 1C as amended

(42 U.S.C. 2133.2134. 220L 32,2233,22=36

2239L 2282) secs. 20 1.0=. 0, W8 Stat. 1=42.

I4. 124. as amended (42 U.SC. 584L S842.

5848); unless otherwise noted.

Section 5U also Issued under Pub. L.S01 sec. 10. 92 Stat. 2M (42 U.S.C. 5851).

Sections MAL58 S0*1 and S032 also issued

under Pub. L 97-415.95 star. 2073 (42 U.S.C.

2239). Section S0.8 also issued under sec.

=22. Stat. W (42 U.S&C. 21.). Sections

50.80.-S081 also Issued under sac. 184 68 Stat.

94. as amended (42 U.S.C 2234). Sections

S0.100-50-102 also issued under sec. 188. 6

Stat. 955 (42 US.C. 223)..

For the purposes of sec. 22 68 Stat. K58. as

amended (42 U.S.C. 223. £160.10 (ae; (b).

and (ci, 50.44.50.48.50.48, S0.54. and 60.W'a)

are Issued under sec. 18b," 68 Stat. 948. as

amended (42 US.C. 2201(b)); 1 50.10 (b) and

(c) and 50.54 are Issued under sec. 161L 68

Stat. 949. as amended (42 U.SC. 2201(i}} and

I5I 550(e). 50.5O.b). 50.n0. 571.50.7. and

50.78 am issued under sec. 1610. 68 Stat 950.

as amended (42 U.S.C. 2=01(o).

L A new J 50.73 is added to read as

follows:

J 50.72 Ucensee event report system.

(a) Reportable events. (1) The holder

ofan operating license fore anuclear

power plant (licensee) shall submit a,

Ucensee Event Report (LER) for any

event of the type described in this

paragraph within 30 days after the

discovery of the event Unless otherwise

specified In this section, the licensee

shall'repart an event regardless of the

plant mode or.power level. and

regardless of the significance of the

structure, system. or component that

initiated the event.

(Z) The licensee shall report-

(I)(A) The completion of any nuclear

plant shutdown required by the plant's

Technical Specifications; or

(B) Any operation or condition

prohibited by the plant's Technical

.pecifications: or

(C) Any deviation from the plant's

Technical Specifications authorized

pursuant to J 50.54(x) of this part.

(11) Any event or condition that

resulted in the condition of the nuclear

power plant. including its principal

safety barriers, being seriously

degraded, or that resulted in the nuclear

power plant being:

JA) In an unanalyzed condition that

significantly compromised plant safety.

(B) In a condition that was outside the

design basis of the plant; or

(C) In a condition not covered by the

plant's operating and emergency

procedures.

(il'] Any natural phenomenon or other

external condition that posed an actual

threat to the safety of the nuclear power

plant or signiiantly hampered site

personnel fix the performance of dutes

necessary for the safe operation of the

nuclear'power plant.

(lv) Any event or condiffon that

resulted in manual or automatic

actuation of any Engineired Safety

Feature (ESF). Including the Reactor

Protection System (RPS). However.

actuation of an ESF. including the RPS.

that resulted from and was part of the

preplanned sequence during testing or

reactor operation need not be reported.

(v) Any event or condition that alone

could have prevented the fulfillment of

the safety function of structures or

systems that are needed to.

(A) Shut down the reactor and

maintain it in a safe shutdown

condition;

(B) Remove residual heat;

(C).Control the release of radioactive

material: or

(D) Mitigate the consequences of an

accident.

(vi) Events covered in paragraph

(a)(Z)(v) of this section may include one

or more procedural errors, equipment

failures, and/or discovery of design.

analysis. fabrication, construction, and/

or procedural Inadequacies. However,

individual component failures need not

be reported pursuant to this paragraph if

redundant equipment In the same

system was operable and available to

perform the required safety function.

(vii) Any event where a single cause

or condition caused at least one

independent train or channel to become

inoperable in multiple systems jr two

independent trains or channels to

become inoperable In a single system

deslgned'to:

(A) Shut down the reactor and

maidtain it An a ale shutdown

condition:

(B) Remove residual heat

(C) Control the release of radioactive

materialh or

(D) Mitigate the consequences of an

accident.

(viii)(A) Any airborne radioactivity

release that exceeded 2 times the

applicable concentrations of the limits

specified in Appendix B, Table B of Part

20 of this chapter in unrestricted areas,

when averaged over a time period of

one hour.

(B) Any liquid effluent release that

exceeded I times the limiting combined

Maximum Permissible Concentration

(PC (see Note I of Appendix B to Part

20 of this chapter) at the point of entrl

into the receiving water (Lea.

unrestricted area) for all radionuclides

except tritium and dissolved noble

gases. when averaged over a time period

of one hour.

(ix) Reports submitted to the

Commission in accordance with

paragraph (a)(2)(vifil of this section also

meet the effluent release reporting

requirements of paragraph 20,405(aXS)

of Part 20 of this chapter.

(x) Any event that posed an. actual

threat to the safety of the nuclear power

plant or significantly hampered site

personnel in the performance of duties

necessary for the safe operation of the

nuclear power plant Including fires.

toxic gas releases, or radioactive

releases.

(b) Contents. The Ucensee Event

Report shall contain:

(1) A brief abstract describing the

major occurrences during the event.

including all component or system

failures that contributed to the event

and significant corrective action taken

or planned to prevent recurrence.

(2)(1) A clear. specific. narrative

description of what occurred so that

knowledgeable readers conversant with

the design of commercial nuclear power

plants. but not familiar with the details

of a particular plant. can understand the

complete event.

(il) The narrative description must

include the following specific

information as appropriate for the

particular event:

(A) Plant operating conditions, before

the event.

(B) Status of structures. components,

or systems that were inoperable at the

start of the event and that contributed to

the event.

(C) Dates and approximate times of

occurrences.

(D) The cause of each component or

system failure or personnel error. if

known.

(E) The failure mode, mechanism, and

effect of each failed component. if

kr,',n.

(F) The Energy industry Identification

System component fumction identifier
Federal Ret'ster I VoL 48. No. 144 I Tuesday. July 28, 1983 / Rul:.,, and Regulations 33859

and system name of each component or

system referred to In the LER.

(1) The Energy Industry Identification

System is defined in: MEE Std 803-1983

(May IS. 1983) Recommended Practices

for Unique Identification Plants and

Related Facilities-Principles and

Definitions.

(2) IEEE Std 803-1983 has been

approved for incorporation by reference

by the Director of the Federal Register.

A notice of any changes made to the

material incorporated by reference will

be published in the Federal Register.

Copies may be obtained from the

Institute of Electrical and Electronics

Engineers, 34s East 47th Street. New

York. NY 10017. A copy Is available for

Inspection and copying for a fee at the

Commission's Public Document Room.

1717 H Street. NW.. Washington. D.C.

and at the Office of the Federal Register.

1100 L St. NW.. Washington. D.C.

(G) For failures of components with

multiple functions, include a list of

systems or secondary functions that

were also affected.

(H) For failure that rendered a train of

a safety system inoperable, an estimate

of the elapsed time from the discovery

of the failure until the train was returned

to service.

(I) The method of discovery of each

component or system failure or

procedural error.

(Ji)() Operator actions that affected

the course of the event. Including

operator errors, procedural deficiences.

or both. that contributed to the event.

(2) For each personnel error, the

licensee shall discuss:

(i) Whether the error was a cognitive

error (e.g., failure to recognize the actual

plant condition, failure to realize which

systems should be functionin& failure to

recognize the true nature of the event) or

a procedural error,

(ii) Whether the error was contrary to

an approved procedure, was a direct

result of an error in an approved

procedure, or was associated with an

activity or task that was not covered by

an approved procedure:

[iii) Any unusual characteristics of the

work location (e.g. heat. noise) that

directly contributed to the error. and

(iv) The type of personnel Involved

Me., contractor personneL utilitylicensed operator, utility nonlicensed

operator, other utility personnel).

(K) Automatically and manually

Initiated safety system responses.

(L) The manufacturer and model

number (or other Identification) of each

component that failed during the event.

(3) An assessment of the safety

consequences and Implications of the

event. This assessment must include the

availability of other systems or

components that could have performed

the same function as the components

and systems that failed during the event.

(4) A description of any cprrective

actions planned as a result of the evez4

including those to reduce the probability

of similar events occurring in the fature.

(5) Reference to any previous similar

events at the same plant that are known

to the licensee.

(8) The name and telephone number of

a person within the licensee's

organization who Is knowledgeable

about the event and can provide

additional information concerning the

event and the plant's characteristics.

(c) Supplemental informanon. The

Commission may require the licensee to

stibmit specific additional information

beyond that required by paragraph (b)

of this section if the Commission finds

that supplemental material Is necessary

for complete understanding of an

unusually complex or significant event.

These requests for supplemental

Information will be made in writing and

the licensee shall submit the requested

information as a supplement.to the

initial LER.

(d) Submision of reports. Licensee

Event Reports must be prepared on

Form NRC 308 and submitted within 30

days of discovery of a reportable event

or situation to the U.S. Nuclear

Regulatory Commission. Document

Control Desk. Washington. D.C. 20555

The licensee shall also submit an

additional copy to the appropriate NRC

Regional Office listed in Appendix A to

Part 73 of this chapter.

(e) Report legibiity. The reports and

copies that licensees are required to

submit to the Commission under the

provisions of this -section must be of

sufficient quality to permit lefible

reproduction and micrographic

processing.

(I) Fxempdona Upon written request

from a.llcenses Including adequate

justification or at the Initiation of the

NRC stafl, the NRC Executive Director

for Operations may. by a letter to the

licensee, grant exemptions to the

reporting requirements under this.

section.

(g) Reportable occurences. The

requirements contained in this section

replace all existing requirements for

licensees to report "Reportable

Occurrences" as defined In individual

plant Technical Specifications.

The following additional amendments

are also made to Parts 20 and 50 of the

regulations In this chapter.

PART 20-STANDARDS FOR

PROTECTION AGAINST RADIATION

3. In 20.402. paragraph (a) is revised;

the introductory text of paragraph (b) is

revised: and a new paragraph (e) is

added to read as follows:

§20.402 'Reports of theft or loss of

licensed materiaL

(a)(1) Each licensee shall report to the

Commission. by telephone. Immediately after it determines that a loss or theft of

licensed material has occurred in such

quantities and under such circumstances

that It appears to the licensee that a

substantial hazard may result to persons

in unrestricted areas.

(2) Reports must be made as follows:

(I) Licensees having an Installed

Emergency Notification System shall

make the reports to the NRC Operations

Center in accordance with j 50.72 of this

chapter.

(H) All other licensees shall make

reports to the Administrator of the

appropriate NRC Regional Office listed

in Appendix D of this part.

(b) Each licensee who makes a report

under paragraph (a) of this section shall.

withing 30 days after learning of the loss

or theft, make a report in writing to the

U.S. Nuclear Regulatory Commission.

Document Control Desk. Washington.

D.C. 20555, with a copy to the

appropriate NRC Regional Office listed

in Appendix D of this part. The report

shall include the following information:

(e) For holders of an operating license

for a nuclear power plant. the events

included in paragraph (b) of this section

must be reported in accordance with the

procedures described in 150.73 fb). (c).

(a). (e)l and (S) of this chapter and must

include the information required in

paragraph (b) of this section. Events

reported in accordance with 1 50.73 of

this chapter need not be reported by a

duplicate report under paragraph (b) of

this section.

4. In 1 20.403. the introductory text of

paragraphs (a) and (b) is revised, and

paragraph (d) ts revised to read as

follows:

S20.403 Nottications of luldents.

(a) Immediate notificaon. Each

licensee shall Immediately report any

events involving byproduct, source, or

special nuclear material possessed by the licensee that may have caused or

threatens to cause:

(b) Twenty-four hour notiicat••on.

Each licensee shall within 24 hours of

discovery of the event. report any event

involving licensed material possessed
33860 Federal R tWw / Vol. 48. No. 144 / Tuesday. lulv 2M. 1983 I Rules and Reulations

by the licensee that may have caused or

threatens to cause:

(d) Reports made by licensees in

response to the requirements of this

section must be made as follows:

(1) Licensees that have an installed

Emergency Notification System shall

make the reports required by paragraphs

(a) and (b) of this section to the NRC

Operations Center in accordance with

§ 50.72 of this chapter.

(2) All other licensees shall make the

reports required by paragraphs (a) and

(b) of this section by telephone and by

telegram. mailgram. or facsimile to the

Administrator of the appropriate NRC

Regional Office listed in Appendix D of

this part.

5. In 120.405, paragraphs (a) and (c)

are revised, and new paragraphs (d) and

(e) are added to read as follows:

S20.405 Reports of ovfwipoair" Wnd

excessive levels and concertt -on&

(a)(1) In addition to any notification

required by I 20.40 of this part. each

licensee shall make a report In writing

concerning any one of the following

types of incidents within 30 days of Its

occurrence-.,

(i) Each exposure of an individual to

radiation in excess of the applicable

limits in If 20.101 or 20.104(a) of this

part. or the licenser.

(ii) Each exposure of an individual to

radioactive material in excess of the

applicable limits in if 20.103(a)(1L),

20.103(a)(2). or 20.104(b) of this part. or

in the license;

(iii) Levels of radiation or

concentrations of radioactive material in

a restricted area in excess of any other

applicable limit in the license;

(iv) Any incident for which

notification is required by 1 20.403 of

this part; or

(v) Levels of radiation or

concentrations of radioactive material

(whether or not involving excessive

exposure of any individual) in an

unrestricted area in excess of ten times

any applicable limit set forth in this part

or in the license.

(Z) Each report required under

paragraph (a)(1) of this section must

describe the extent of exposure of

individuals to radiation or to radioactive

material. including:

(I) Estimates of each individual's

exposure as required by paragraph (b)

of this section:

(1i) Levels of radiation and

concentrations of radioactive material

involved-

(iii) The cause of the exposure, levels

or concentrations; and

(iv) Corrective steps taken or planned

to prevent a recurrence.

(c)(1) In addition to any notification

required by 1 20.403 of this part. each

licensee shall make a report in writing of

levels of radiation or releases of

radioactive material in excess of limits

specified by 40 CFR Part 190.

"Environmental Radiation Protection

Standards for Nuclear Power

Operations." or in excess of license

conditions related to compliance with 40

CFR Part 190.

(2) Each report submitted under

paragraph (c)(t) of this section must

describe.

(i) The extent of exposure of

individuals to radiation or to radioactive

material;

({i) Levels of radiation and

concentrations of radioactive material

involved:

(Of) The cause of the exposure, levels,

or concentrations: and

(iv) Corrective steps taken or planned

to assure against a recurrence, including

the schedule for achieving conformance

with 40 CFR Part 190 and with

associated license conditions.

(d) For holders of an operating license

for a nuclear power plant, the incidents

included in paragraphs (a) or (c) of this

section must be reported in accordance

with the procedures described in

paragraphs 50.73 (b). (c), (d). (el, and (g)

of this chapter and must also include the

information required by paragraphs (a)

and (c) of this section. incidents

reported in accordance with 1 50.73 of

this chapter need not be reported by a

duplicate report under paragraphs (a) or

(c) of this section.

(e) AlU other licensees who make

reports under paragraphs (a) or (c) of

this section shall, within 30 days after

learning of the overexposure or

excessive level or concentration, make a

report in writing to the U.S. Nuclear

Regulatory Commission. Document

Control Desk. Washington, D.C. 20555,

with a copy to the appropriate NRC

Regional Office listed in Appendix D of

this part.

PART 60--DOMESTIC LICENSING OF

PRODUCTION AND UTILIZATION

FACILITIES

6. In I 5038 new paragraphs (c)(6)

and (7) are added to read as follows:

f50.36 Technical specicatons.

(c)

(6) initial Noti7cation. Reports made

to the Commission by licensees in

response to the requirements of this

section must be made as follows.

(I) Licensees that have an installed

Emergency Notification System shall

make the initial notification to the NRC

Operations Center in accordance with

1 50.72 of this part.

(ii) All other licensees shall make the

initial notification by telephone to the

Administrator of the appropriate NRC

Regional Office listed in Appendix D.

Part 20. of this chapter.

(7) Written reports. Holders of an

operating license'for a nuclear powerplant shall submit a written report to the

Commission concerning the incidents

included in paragraphs (c) (1) and (2) of

this section in accordance with the

procedures described in 1 50.73 (b). (c),

(d), (e). and (g) of this part. Incidents

reported in accordance with 150.73 of

this part need not also be reported under

paragraphs (c)'(1) or (2) of this section.

Dated at Washingto D.C this 20th day of

Julyi9

For the Nuclear Regulatory Commission.

Samuel J. Chkm

Secretary of the Cmmission.

IFM Dor- 0-MIN W-id 74-ft MS =m

eLtLINQ CODE 758041

APPENDIX F

1992 REVISION TO 10 CFR 50.72 AND 50.73 INCLUDING

STATEMENT OF CONSIDERATIONS

Published in the Federal Register

on September 9, 1992 September 10, 1992

(Vol. 57, No. 176, pages 41378-41381)

This Federal Register notice does not provide a complete version of 10 CFR

50.72 and 50.73; it addresses only small parts of those sections. Its purpose

here is to present the Statement of Considerations, which explains some of the

reporting requirements of the sections.

NOTE:

NUREG-1022, Rev. I

41378 _&W~g~t I VoL 57~.Po. up7I. Thtruday. seotember0,-o "n9 / Rules ilad Regulatlp=

certified to OMB. In a letter dated

August 14.1992. that by unanimous vote

the Commission had overridden the

OMB'1 disapproval of the information

collection request associated with this

rule.

On August 21.1992. OMB assigned.the

following new control number:. 3150-

0171. effective until August 31; 1995.

This new control number Is only

applicable to the sections in 10 CFR part

35 amended by this rule. Information

collection authority for all other sections

of 10 CFR part 35 remains under the

existing general control number: 3150-

0010.

Ustof Subjects in 10 CFR Part 3S

Byproduct material. Criminal penalty.

Drugs. Health facilities. Health

professions. Incorporation by reference.

Medical devices. Nuclear materials&

Occupational safety.and health.

Radiation protection. Reporting and

recordkeeping requirements.

Text of Final Regulations

For the reasons set out in the

preamble and under the. authority of the

Atomic Energy Act of 1954. as amende&

the Energy Reorganization Act of 1974.

as amended, and S U.S.C.552 and=53,

theNRO Isaddpting the fbllowing iznendments to 10 C part 35.

PART 35-MEDICAL USE OF

BYPROOUCTMATERIAL

1. The authority t.uation for part continues to read in part as follows:

Autidtyr. Sacs; 1eL estaL 94L as

amended (42 U.&.C. l)k sec. m01. a Stat.

142.4s'amended44ZU.rC.5841l" * 0.

2. In I 35.. paragraph (b) Is revised

und paragraph (d) Is Wdded to read as

follows:.

1f 3U iformatton coollction

requirementw: MB appovaL.

(b) The approved information

collection requirements contained in this

part appear in &1 =5.1.35.13. 35.14.

35.21. 3522. 35.2&35.27. 35.29,&5.3L

350 35.5L 35.53.35..3.00.35AL.6

35.=0 35.8.35.2. 35.204 35.=0535Z310.

S53215.35)4K 35.40&. I5I4i0, 35.41S.

35.606. 35.810.35.615. 35.630. 35.03.

35M 4.35.3 35.-4L 35.643. 35.545. and

35.47.

(d) DMB has assigned control number

3150-0171 for the information collection

requirements contained in I 1353 and

35.33

Dated at Rockville. Maryland. tntv3d day of September 1992.

For the Nuclear Regulatory Commisuion.

Samuel I. CIlk.

Secetary of the Commission.

(FR Doc. S2-31754 Filed 9 0 Z- &:45 amm

B1.I40 COOC 694-t.M

10CFR Part 50

RIN 3150-AE12

Minor Modifications to Nuclear Power

Reactor Event Reporting

Requirements

AGENCr. Nuclear Regulatory

Commission.

ACTiOI: Final rule.

SWUARr, The Nuclear Regulatory

Commission (NRC) has amended its

regulations to make minor modifications

to the cirrent nuclear power reactor

event reporting requirements. The final

rule applies to all nuclear power reactor

licensees and deletes reporting

requirements -for some events that have

been determined toý be of little or no

safety significance. The final rule

reduces the industry's reporting burden

and the NRCs'response burden in event

review and assessment.

EFFEcmT CoATE: October 13. 129

FOR FURTHER NVORIMTION CONTACTI'

.Raj4 T"tfath. Oflfe for Analysis and

Eviluation of'Operadonal Data. U.S.

Nucl.ear qgulatory.Cormnlsslon.

&UPPLEMENARY U(FORMAT01OC

Backgound

The Commission is issuing a final rule

that amends the nuclear, power reactor

event reporting requirements contained

In 10 CFR 50 .72 'Immediate Notification

Requirements for Operating Nuclear

Power Reactors." and 10 CFR 5.3.

"Licensee Event Report System." The

final rule is issued as part of the

Commiselon's ongoing activities to

improve its regulations. Specifically. this

final rule amends 1o CFR s5. (b)C2)(1)

and 160-R S.7 (a)(2)(iv). On June 28.

1992 (57.ER. 254M2). the Commission

issued a proposed rule requesting public

comments on these amendments.

Over-the past several years. the NRC

has Increased its attention to event

reporting issues to ensure uniformity.

consistency, and completeness in

reportn8g4n Sepfeinber 19 the NRC'a

Office forAnulysis and Evaluation of

Operational Data (AEOD) Issued for

comment a draft NUREG-.,022 Revislo,

1.1 "Event Reporting Systems 10 CFR

I freesh*e apy may be nquaeted by writdn to

dw DLstztbw utia n Madl Service. Section. US.

5.2 and 0-.CFR 50j..C~arification of

NRC Systems and Giddelines For

Reporting." Following resolution of

public comments. the .NUREG will be

issued In the final form. The NUREG

will contain improved guidance for

event reporting.

NRC's reviews of operating

experience and the patterns of licensees'

reporting of operating events since 1984

have indicated that reports on some of

these events are not necessary for the

NRC to perform its safety mission and

that continued reporting of these events

would not contribute.useful information

to the operating reactor events

database. Additionally., these

unnecessary reports would have

continued to consume both the

licensees' and the NRC'6 resources that

could be better applied elsewhere. The

NRC has determined that certain types

of events..prlmarilythose involving

Invalid engineered sfetyfeilure(ESF)

actuations. are of littleor.no.safety

significance.

Valid ESF actuations are mose

actuations that result from .valid

signals" or from intentional mauual

initiation. unless It Is partof

preplanned test. Valid signals,.arethose

signals that aurpe natadi ,re.spuse to

actualpPlant c6nditio6.s6-r-pAemters

satisfying the reqeeTsoriESF inltia'"n+

invalidactuationsare-bdefinitiozi

those thai-do notateet t6e. raor

being valid.Thus. invalidactuations

Include actuations that'are-notthe result

of valid signals and are not Intentional

manual actustions..Invalid actuations

include instances where instrument

drift. spurious signals. h6man error. or

other invalid signals caused actuation of

the ESF (eg. jarrig #.cabinet. an error

in use of Jumpers of lifted ieads an error

In actuation of switches or controls.

equipment fallure..or radio frequency

interference).

NRC's evaluation of both the reported

events since January 1U4. when the

existing rules'first became effectiveý and

the comments received during the"Event.

Reporting Woi'kshops conducted In Fall

of 1990 identified hec4e4ldeowyements.

in the rules *he'NRC dE.rrmhxe8 96L

invalld actuatdon.. solation, or

realignment of a limited-set of ESFs

including the systems. subsystems. or

components ji~e.. an invalid actuation.

isolation. or realignment of only the

reactor water dean-up-PItWCUJ aystem.

Nuclear Replatoy Comw-u-•aa-.Wushhquoa.DC

ZW8. A copy to also avail"l for hopectim cc cpyt 6or a he it the NRC Public Downt

IRom. Z20 Llret. NW.ULower Levil.-

Washltcstoi DC =6&
Federd-ReOlw4g, W

R41-

the controtroom emeeic.y:.entiat•o•-

(CREv) system, the reactor "Ibulding ventilationjsystem. the fuel buildm; ventilation system, or the auxiliary

building ventilation system, or their

equivalent ventilation systems] are of

little or no safety significance. However.

these events are currently reportable" under 10 CFR 50.72 (b)(2)(Ui) and 10 COR

5. (a)(2)(lv) , The final rules for the current event

reporting regulations. 10 CFR 50.72 and

10 ClR 50.73 (48 FR 39039M August 29.

1983. and 48 FR 338.40 July 20.1=83,

respectively). statedthat ESF systems. including the reactor protectiofh system (MPS). are provided to mitiate the

'consequences of a significanit event.

Therefore. ESFs should'(1) work

properly when called upon and (2) should not be chailenged'frequendy or

unnecessarily. The Statements of

Consideration for these final rules.also

stated that operation of an.ES? as-ppa.• of a pro-plannmd operational pr•cedure

or test-need not be reported. The

Coummission noted ht E ectuauons.

including reactor trips, are frequently associated with si~niflcantplant transients and a reindicative of events.

that are-of safety.significance. At that

time. the Commission also required al

ES? actuations.including theR.PS

actuations, whether manual-or

automatic.valld or-Invalld-except as

noted.10 be reported to the.NRC by- telephone within 4 hours of occurrence

followed by a writte aLcensee event

Report (LMI) withln30 days of the

incident. Thit requirement on.timeliness

ofreporting remalns unchanged The'reported information Is.used by the NRC In confirmation.of the licensing bases, identification of procursors-to

severe core damage..idehtilcatlon of

plant specific deficlenciesa.eneric

lessons, review of management control

systems, and licensee performance

assessment.

Discussion

The NRC has determined that some

events that Involve only Invalid ES?

actuations are of little orno safety

significance. However, not ea Invalid

ESF actuatious are being exempted from

reporting through this rue. The

relaxations in event reporting

requirements contained in the final rule

apply only to.a narrow, limited set of

specifically defined invalid ES?

actuations. These events include Invalid

Atation. isolation, or reaHgnment of a

Pm.tedset of EF ihnc•ding systems isyibsystems,-or components.(.e., an

Invalid actuation. Isolatiom or

realignment of only the.RWCU.wstem.

or thecm system Teactor buildMS

ventilation systemuelbulding

venttlatiwosystem-,auxtliarybuilding

ventatton-sy•sietn ielrýequ ezit

ventilatfon'systems .The :tutiion of

the'standby gas treatment system following an invalidactuation of the

reactor building ventilation system is

also exempted from reporting. In

addition, the final rule excludes •-valid

actuations of these ESFa-or their

equivalent systemst.from sinals that ornowed 5rom nm*-ESF circu-itry,

However. invalid ictuatlons of other

i.Fs would continue to be reportable.

For example Jmnergency.core coolinz

system isolations/actuationx.

containment -isolatiwn-valve closures

that affect coolg ,yiteis. main steam

low, e#sential iippdt systeims, etc..-

containment spray.actuaton: -and

residual beat removalisystem isolations

(orisystems deslgnhted.by-any other

names but designed-to fulfil) the

function-similar to.th'es systems. and

their equivalents). 'ar sill Ieportable..If

an invalid Esiectuatiow'eveils a

defect in thesystem so thatthe system failed or would-fall.to permorm its

intended fuiction, the eii& contlnues to

be rep bli under'bier reqirements of 10 CFR 50.7 and:1O CFR 50,.13f a

condition or deficiency has (U an

adverse impact an "letyrelated

equipment and consequently on the

ability to shut down the reactor and

maintain it nea sale'shutdown

•condition, (Z) has &pjotential for

signifiat radntogfcelass 6r

potential expo5Us1 toplant personnel or

the general public, or (3)'would

compromise control room habitabillty. the event/dfscovery continues to-be.

reportable,

Invld ES actuations that are.

excluded by.this fial rule,.but occur as

a part of a reportable ve.nt, continue to

be describe4 as partof the reportable

event. These amendcents 'not ar 

intended to.preclude.submittal of a

complete, accurate, and.thorou

description of an event that Is otherwise

reportable under-.0 CFR.50.7Z or 10 CFR

50.73. The Commission relaxed only the

selected event reporting requirements

specified In this final rule.

Licensees ar still required under 10

CFR part 50O appendix %,"Quality

Assurance-Criteria forN~nb•le Power

Plants and Fuel-Reprocessing Plants.".to

address corrective actions for events or

conditions that 'are adverse to quqilty whether the event Is reportable or not.

In addition..imlnlmi-ing• EF 8ttuations

(avch asRWCU Isolations) to reduce

operation rad litlon exposures

assodited With the lnvesltimton and

recovery.from the. jtitons, are

constrent with ALARA-requirements.

events ro

(1).ThU first'.itegry excludes events

in which an invalid ESPor RPS

actuation occurs when'the.system is

already properly remoiedofrm service

if all requirementsof plant prcedures for removing equipment from service

have been met. This includes required

clearance documentation, equipment and control board tagging.'andproperly

positioned valves and power-supply

breakers

(2) The second category excludes

eventrini.,vhlchiulinvalid ESP.br.RPS

actuation occurs after the safety functionhaa alreadybeenzdampleted.

(e~g., an-inv'alid contklnment'iolatlon.

signal while the containment Isolation-

-valves aiv already clo6ed.6.an Invalid

actuation, of [he RMS wheaUll rodsaete

filInserted).

I 3) Thetdirid ratetgy.exludes even

In which An invalid ESP1 ackuatlon

containment spray aactaflou; residual

hema removal systm isolaatm. thieir equivalent systems). •toninue to be

rep rtatle.

I e continue to be.required to

subiuit LERsIfa deficency or condition

associated with any of the i•nilid ESF

actuations of the RWCU of the CREV

systems (or other equivalent ventilation

systems] satisfies anyor bllity criteria under I50.72 1 50M."

Impact of the Amendments the

Industry and Government Resources

Relaxing the requirement for reporting of certain types of ESPActuatlons

reduces the industry' reporting burden

and the NRC response bui.en. This

reduction is consistent with the

objectives and the requirdnitnts of the

Paperwork Reducti•n Act, Use

amendi'ents have no Impact on-the

-NRG's ability to A I M ml .fW to

ensurepublio health dnd fe•ty.because theideletedieportabilfty reqnlents have little ar no safety ululi_ mnc1

It is-estimated- that the- .toalhe

exitinzg zes wi i-sultl ibout-5l0 for
in "qto~f 'v~aa - -. -~ -- I - r ~

Reports each year. Shnl•

are expected In the numnberfrompt

event notifications repoit e mideria

CFR 50.7. Some respondentsAntheir

comments on the proposedtrule.dated

June 28.1992. submitted an estimate of

approximately is percent reduction in

their reporting burden.

Summary of Comments

The NTRCeceved 20 comments--Z

from individuals. 3 from industry.

supportedcrganhatims.-and14 from

utillties.tceptfr twozespondentsadl

commenters welcamedthe

Commits efforts torednce the

licensee burdenAnd to save the

agency's esoures in event eeview and

processing The utilities mnd he

industry-upportedorganhations

expressed their desire for a broader

relaxation to. dd invalidESd

actuation•ina~epoAt

Other fi=6azeepoades

concerned the tlloit-carficationf

the dehtiUn•nva d",Voas

examples of evnts belqS eempted foivii reprtcg~demiowpfVIWlnt specaatlodt s:,xemption frM

reporth of-heaiatioof thestandby

gas tratnt systenish waa

Invaliduatuation o h ratrbldn ventilation isposib •lly

exte rels •of Invalid.

actuations/lsoalons.0faWLXJ from

Mpxoft to innideýbose afthe chemicalandwol/aie pontrdaystem ia

a pressurized oaterzeactor The

Statement of Copiidereationior this

final ule a desses iostoafhse

concems. Othefr]uet anddcalafcations

concerning e potsbfllrwill be

addressed In NUREZ-I122. Relsonl.

However. It is not practicalo -adds a

plant-specific sitatloi unless it relates

to a generic coecem

The Commission stresses that only

certain specific Invalid ESF actuations

are being exempted from reporting

through the present amendments.

NUREG-z, Revision I Will contain

specific exampe and additional

guidance on events which ar presently

reportable as well as those which are

being exempted from teporting through

these amendments. Inthe future, the

Commission will give due consideration

to other proposed relaxations from

event reporting after the NRC staff has

had an opportunity to reassess the data

needs of the agencyand pe donie

safety uses$mmids to lusti ntiating

separate getieraLrdlerma11 .Until such

time. aUl' venti notzp exempted in * these .s connue

to be reportable. The wo resndents who opposed

the proposed amendments expressed

theiatcinerns 4bout minating the

seleiled event reporting requirements.

These comnmenters believe that the

eliminatonmf these event-reporting requirements may adversely affect the

NRCs information database -and

ultimately effect the agency's ability to

carryout its mission to protect public

health and saety. For many years. the

NRC staff has been systematically.

reviewing information obtained from

Licensee Event Reports. These

assessments of reactor operational

experience have Icinded data on the

types of events Included Inthe three

categoriesthatthe oNRC is deleting from

reporting. The staffs reviews and

assessments of nearly 1000 reactoryears of operational experience have

Identified esentially no safety sgnficae associated Nith the type"Of

events Included in the aoementioned

three categores. The Commisslonhas

reviewed the scope ofthese

amendment'ma n the basis of the

staff s assessment of the past eactor

operational experience" has

subseuentl ocladed svith a

reasoale confidence that relaxation

from reporting of events In the three

categories does mot affect the agncy's ability t6o ntectpublibealth and

safety. . Based on the Input f•om the utilities.

these amendments wil redue the

Industry's reporting burden by-about 15

percent. The estimated savings of the.

NRC's response burden'In event revIew

and assessment Is about 5-10 percent.

ont•I mpact: Categaoical

Exclusion

The NRC has determined.that this

final rule is .the type of action descibed

In categorical exclusions 10,CF GZ

(413M) and (14}. Therefore. neither an

envircnmental Impact statement noron

environmental assessment has been

prepared f(r this final rule.

Paperwork Reduction Act Statement

This finalrale amends information

collection requirements that are subject

to the Paperwork Reduction Act of 1980

(44 UQ.C. 3501 et seq.). These

amendments were approved by the

Office of Management and•fludget approval numbers 310-00m1 and 3150-

Because the rule will relax existing

reporting requirements. public reporting

burden of information is expected to-be

reducedi.It Is estimated that about 10

fewer Licensee Event.Reports (NRC

Form S86) and a sinmlarl reduced

number of'prompt event notifications.

made pursuant tol10 CFR W-.72 wIll be

required each year.Tke tsulting reduction In burden Is estimated to

average T0 hours per licensee response.

Including the time required rviewing

Instructions. searching existing data

sources, ga henngn lntaining the

data needed. andAeviewing the

collection of information. Send

comments regarding the estimated

burden reduction -or any other aspect of

this collection of information. including

suggestions for seducing this burden, to

the Information and Records

Management Branch (MNBB-7M4). U.S.

Nuclear Regulatory Commission.

Washington. DC2O5s and to the Desk

Officer. Offic of infornation and

Regulatory Affars tNOB-3MO. (3150-

0011 and 3150-0104),Offlce of

Management and Budget. Washington.

DC Zeto

Regulatory Analydsi

The Commlaslonaas prepared a

reg4atory analysis an is final tule.

The analysis examines the costs and

benefits of he -altematives msIdered

bythe Comnumssion. The analysis is

available for Inspection In the NRC

Public DocumeutRoon.2 IZLStreet.

NW. lower Level, Washigton. DC

=Zoss.Single copies of the analysis may

be obtained firm: RajlTdpafL Office

for Analysls and Evaluation of

Operational at USI Nuclear

RegulatoryCommiUss•. WashingtoM.

DC 20555. Telephone 3013) 492-4435.

Regulator FlexibilltyOUICorifaion

In accordance with the Reylatory

Flexibility Act of 120 (5 IS.C. W0 (B)).

the Comnmssion certifies that this.rule

does not haves sgni icant economic

impact on a substantil number of small

entities.-The final rule affects only the

event reporting requirements for

operational nuclear poier plants. The

companies thatvwnthese plants do not

fall within the scope of-the definition of

"small entities" set forth In the

Regulatory Flexibility Act or the Small

Business Size Standards set out In

regulations Issued by the Small Business

Administration Act In 123 CFR part 12t.

Backfilt Analysis

As required by 10 CFR S0.109. the

Commission has cdmpleted an

assessment of the need for Bacidit

Analysis for ibis final rule. The

proposed amendments Include

relaxations of certain existing

requirements onzeportiq ofinformatton

to -the NRC.These changes neither

impose additional reporting

requirements. nor adqufre modifications

to the facliti or.their lc"enseL

Accordingl l doe.Chas concluded

that this final rule does not constitute a
Fedeal RegisWer VoL-57-..INo. 178 _Th ada pebv09. Rtl degjs418

backfit and. thus. a-backflt analysis is

not required.

List of Subjects in 10 CFR part 50

Antitrust. Classified information.

Criminal penalty. Fire prevention.

Incorporation by reference,

Intergovernmental relations. Nuclear

power plants and reactors, Radiation

protection. Reactor siting criteria.

Reporting and recordkeeping.

For the reasons set out in.the

preamble and under the authority of the

Atomic Energy Act of 1964. as amended.

the Energy Reorganization Act of 1974.

as-amended. and 5 U.S.C. 532 and 553,

the Commission is adopting the

following amendments to-1O C1R part 5O.

PART 50--DOMESTIC LICENSING OF

PRODUCTION AND UTILIZATION

FACILITIES

1. 'The autnority citation for Part 50 Is

revisid to read as follows:

Authodat. Secs. 20Z .103. 04.10. 26s. az.

83.18. 18& . 68 Stat. 935.937. 93W.& 4.

954.9 M.0A as amended. sc. Z34. 83 Stat.

t4. as amended 142 U.S.C C =2133.2134.

2135.2201.2212.2233. .23L2.2M 2282Uý sgecs 201. as amended. 202.205. 88 StaL 2242. as

amended. 1244.1248 (42 U.S.C. 5414

584). Section 50W se Issued under Pub. L 9S0.-smec 10. i2 Stat 2•0 (42 U.S8S.• i

Section 0.10 also Issued under secs. 10n 1i5

6 Stat; M M& as amended (42 UA.C. 2131

2=s) s.mtoz. Pub.L 31-190 83 Stat 153 (42 U.S.C 4332) Sections 80.1 3. 04(dd and W0.103 also Issued under ee..10. Go Stat 939.

as amended (41U.SC 28). Secton. 50.=, 50.3, 50.45, and S.5 also Issed under sac.

18• 68 Sta 9W4 (42 U.S.C 2235) Sections

0.a. 50 SU and Appendix Q also issued

under se 102. Pub, L 1-1o 83 Stat 853 (4 U.S.C. 4332) Sections W0.4 and 50.54 also

Issued under sec. 24. 88 Stat,.1245 (42 U.S.C.

5844) Sections 5.153&05U2. and 5An9 also Issued under Pub6 L 07-41. 96 Sta 2073 (42 U.S. 9 Section 58alsoh Issuedunder

se. 122. 8 Stat. 939 (42 U.SC. 2154 Sections

5004. also Issued under se 164.6 Stat

954 as amended (42 U.S.C. 2234). Appendix F

also issued under se. 18. 8 Stat. 55 (42

U.S.C. 2237). For the purposes seO 63 Stat 9. as

amended (42 U.S.C. 227 3t I 5".5o.4e(s) and (b). and S0.S4(c) are Issued under sac.

18b. 68 Stat 94M as amended (41 U.S.C.

=21(b)): 1.3.5.I,ý7), 0 5o.1(asc). 50.34(s) and (e). WA.44(a-c). W56(a) and (), 547(b).

W046(a) KC (d). and K 35&.49(a); 50.54a) 01) (IXM). ")-(n). (pA. (q). (K. Kv). and (y)..504s(. WoSa(s(). ()ce)e (. and (h). 50.5c). 50•.(,a). 50.2(b 50.q4b). 5u8, and s0.8o(a) and (b) am Issued under sM 1611; 08 Stat

949. as amended (42 U.=.C Z21(1): and

111500(d). Orf. and% X4w).jz).-(bb). (cc). and Idd) 80.55e) 50.Wb.59040.6(b 50.8(b) 50470%4) 50.7(a-c) and (e). 50.72(aj-50.73a) and (b). 80.74.75.• and 50.90 are ssued.

under.sc, 1620;.'8 Stat 90% as amended (42 U.S.C 22M(o)).

2. In 1 50.M2 paragraph (b)(2)(i) -1s

revised to read as follows:

15W.72 Inmedlate nottfcstion

requiements $or operating nuclar power

reactors.

(b) Non-emergency Events."

(2) Four-hour reports. * *

(ii) Any event or condition that results

in a manual or automatic actuation of

any engineered safety feature (ESF).

including the reactor protection system

(RPS), except when:

(A) The actuation results from and is

part of a pro-planned sequence during

testing.or reactor operation:

(B) The actuation is invalid and:

(1) Occurs while the system is

properly removed from service:

(2) Occurs after the safety function

has been already completed.or

(3) Involves only the following specific

ESFs or their equivalent systems:

(i Reactor water clean-up system;

(i11 Control room emergency

ventilation system:

(fil Reactor building ventilation

system; (iv) Fuel building ventilation system;

or

(v) Auxiliary building ventilation

system.

3. In,.I 0.7. paragrapha)(2Xlv) is

revised to read as follows:

50'72 " Ucen-e evet repor syster.

(a) Reportable events.& *

(2)-Thlicenseehall report•

(Lv) Any event..or~ondition that

resulted in a manual or automatic

actuation of any engineered safety

feature(ESI,, IncludInthe reactor

protection system {R1% except when:

(A) The actuation resulted from- and

was part of a pro-planned sequence

during testing or reactor operation

(B) The actuation was Invalid and:

(1) Occurred while the system was

properly removed from service:

(2) Occurred after the safety function

had been already completed; or

(3) Involved only the following

specific ESFA or their equivalent

systems:

(i) Reactor water cleanup system;

(th) Control room-emergency

ventilation.system:

(iiM) Reactor building ventilation

system:

(iv) Fuel bu•idingventilatfon system;

or

(v) Auxiliary buildninventilation

system.

Dated at Rockville. MID. this Vth dayof

August1992.

For the Nuclear Regulatmy Commission.

James K Taylor.

Executive Mrectarfor Oveaidons.

(FR Doc. 92-21750 Filed "4k% 8:45 aml

XLWQ CO0K 7mI.41-U

FEDERAL RESERVE SYSTEM

12 CFR Part 225

IRegtdaton Y; Docket No. R-0706)

RIM 7100-AB09

Bank Holding Companes ad Change

In Bank Control

AGEcr:. Board of Governors of the

Federal Reserve System.

ACTnC Final rule.

suMAnu. Thn Board Is.amending Its

Regulation Y to augment the lst of

permissible nonbanking activities for

bank holding companies to Include the

provision of full service securities

brokerage under certain conditions and

the provision of financial advisory

services under €artaln conditions. The

Board has by order previously approved

these activities. Applications by bank

holding• companies to en" In

activities included on the igulafon.Y

list ofpermissible IobAidng activities

under bya theb m ak •.-dtie i s.tn=aluant to

delegae authority.

EFFECTM CATZt SeptemberiO.1092.

FOR FRJ M OWFOMAM CONTACT.

Scott Q. Alvar Associate GeneralCouxsel (/452-35•83a.or Thomas X

Corel. Senior Attorney (20453-M.

Legal Dfvisacln For the hearing Impaired

only. TelecomU•nIcations Device for

the Deaf (MDD). Dorothea Thompson (20214,5,-3544)..

SUPPLEUM AR IN•MATIOI

Background

The Bank Holding Company Act of

1955, as amended (the "lC Act"J.

generally prohibits a banA'holding

company from engaging in noa•nking

activities or acquiring voting securities

of any companyihat it not a bank.

Section 4(c)(8) of the BHC Act provides

an exception to this prohlbitf6n where.

the Board determinew-aftef notice andopportunity for hearing that th.

activities being conductedare "so

closely related to banking'o managing

or cbntrolllft bank. as to be propier

incident thereto-" 12 US.C.1843(cX6).

The"B~ard U atithorizedfl6 make this

determination by order in an indivimlu

case orby regulation.

The Boards Regulation Y (12 CFR part

225) sets forth a list of nonbanking
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(2-89) (Assigned by NRC. Add Vol., Supp., Rev..

NRCM 110.Z and Addendum Numbers. If any.)

320, 3202 BIBLIOGRAPHIC DATA SHEET

(See ins frctions on the reverse)

2. TITLE AND SUBTITLE NUREG-1022, Rev. 1

Event Reporting Guidelines

10 CFR 50.72 and 50.73 3. DATE REPORT PUBLISHED

Second Draft for Comment MONTH YEAR

January 1998

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

D.P. Allison, M.R. Harper, W.R. Jones, J.B. MacKinnon, S.S. Sandin Regulatory

7. PERIOD COVERED enclusive Dates)

8. PERFORMING ORGANIZATION -NAME AND ADDRESS (f NRC, provide Division, Ofte or Region, U.S. Nuclear Regulatory Commission, and mailing address; if conhactr,

provide name and mailing address)

Office for Analysis and Evaluation of Operational Data

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (f NRC. type "Same as above' if contractor provide NRC Division, OTffce or Region. U.S. Nuclear Regulatory Commission,

and mailing address.)

Same as above.

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

Revision 1 to NUREG-1022 clarifies the immediate notification requirements of Title 10 of the Code of Federal Regulations, Part

50, Section 50.72 (10 CFR 50.72), and the 30-day written licensee event report (LER) requirements of 10 CFR 50.73 for nuclear

power plants. This revision was initiated to improve the reporting guidelines related to 10 CFR 50.72 and 50.73 and to consolidate

these guidelines into a single reference document A first draft of this document was noticed for public comment in the Federal

Register on October 7, 1991 (56 FR 50598). A second draft was noticed for comment in the Federal Register on February 7,

1994 (59 FR 5614). This document updates and supersedes NUREG-1022 and its Supplements I and 2 (published in

September 1983, February 1984, and September 1985, respectively). It does not change the reporting requirements of 10 CFR

50.72 and 50.73.

12. KEY WORDS/DESCRIPTORS (L ist words orphrases that will assist resarchers in locating the report.) 13. AVAILABILTY STATEMENT

Emergency Notification System unlimited

Event Report 14. SECURITY CLASSIFICATION

Immediate Notification (Tis Page)

Licensee Event Report unclassified

Notification (this Report)

Report unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2.89)


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WASHINGTON, DC 2055-,9W1

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