Text

Event Reporting Guidelines 10CFR50.72 and 50.73.Second Draft for Comment
	ML20063H691
Person / Time
Issue date:	02/28/1994
From:	Allison D, Harper M, Israel S, William Jones, Mackinnon J, Sandin S; NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:	;
References
	NUREG-1022, NUREG-1022-2DRF, NUREG-1022-R01-DR-FC, NUREG-1022-R1-DR-FC, NUDOCS 9402220125
	Download: ML20063H691 (182)
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i NUREG-1022 i

Rev.1 l

Second Draft i

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! Event Reporting Guidelines 10 CFR 50.72 and 50.73 1

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Second Draft for Comment l

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AVAILABILITY NOTICE f

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Availabihty of Reference Materials Cited in NRC Publications Most documents cited in NRC pubhcations will be available from one of the following sources; f

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The NRC Pubhc Document Room, 2120 L Street. NW., Lower Level, Washington, DC i

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The Superintendent of Documents, U.S. Government Printing Office, Mail Stop SSOP, l

Washington, DC 20402-9328 3.

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NUREG-1022 Rev.1 Second Draft Event Reporting Guidelines 10 CFR 50.72 and 50.73 i

Second Draft for Comment 4

4 Manuscript Completed: February 1994 l

i Date Published: February 1994 1

D. P. Allison, M. R. Harper, S. Israel, W.,R. Jones, J. B. MacKinnon, S. Sandin i

i Office for Analysis and Evaluation of Operational Data l

U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 J

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I ABSTRACT t

Revision 1 to NUREG-1022 clarifies.the immediate notification' requirements.of 1

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 i

the reporting guidelines related to 10 CFR 50.72 and 50.73~and to consolidates i

these guidelines into a single reference document. - A first draft'of. this-j l

document was noticed for public comment in the Federal Register on October 7, 1

1991 (56 FR 50598). 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 i

50.72 and 50.73.

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CONTENTS Page ABSTRACT.................................................................

iii CONTENTS................................................................

v FOREWORD.................................................................. ix EXECUTIVE

SUMMARY

xi ACKNOWLEDGMENTS......................................................... xiii ABBREVIATIONS...........................................................

xv 1

INTRODUCTION.......................................................

I 1.1 Background...................................................

I 1.2 Reporting Guidelines and Industry Experience.................

2 1.3 Revi sed Reporti ng Gui del i nes.................................

3 1.4 How to Use These Guidelines.................................

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REPORTING AREAS WARRANTING SPECIAL MENTI 0N.........................

11 2.1 Engineering Judgment.........................................

11 2.2 Differences in Tense Between 10 CFR 50.72 and 50.73....................................................

11 2.3 Reporting hultiple Failures and Related Events...............

11 2.4 Deficiencies Discovered During Design-Bases Documentation Reviews, Safety System Functional Inspections, and Other Licensee !cgineering Reviews.................................

12 2.5 Engi neered Safety Features Actuati on.........................

12 2.6 Events and Conditions Initially Discussed with the NRC Staff or Identi fied by NRC Inspections...................

12 2.7 Mul ti pl e Component Fai l ures.................................

13 2.8 Human Pe r fo rma nc e I s s u e s.....................................

14 2.9 Voluntary Reporting..........................................

15 2.10 Retraction / Cancellation of Event Reports.....................

15 2.11 Ti me Li mi t s fo r Reporti ng...................................

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SPECIFIC REPORTING GUIDELINES......................................

18 3.1 10 CFR 50.72 and 50.73 General Requirements..................

19 3.1.1 10 CFR 50.72 Immediate Notification Requirements for OperatingNuclearReactors$50.72(a).........................

19 3.1.2 lu CFR 50.73 Licensee Event Report Sys t em 9 50. 73 (a ) ( 1 )..........................................

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3.2 One-Hour ENS Noti fications and 30-Day LER Reports.............

23 3.2.1 Plant Shutdown Required by Technical Specifications

$ 50.72(b) (1) (i) (A) and 650.73 (a) (2) (1) (A).....................

24 3.2.2 Technical Specification Prohibited Operation or Condition 9 50. 73 ( a) ( 2) ( 1 ) ( B)............................................

27 3.2.3 Technical S ecification Deviation per 650.54( ) 650.72(b)(1)(1)(B) and 5 5 0. 73 (a ) ( 2 ( t ) (C )............................................

33 3.2.4 Operating Plant in a Degraded or Unanalyzed Condition 650.72(b)(1)(1i) and 6 50. 73 (a ) ( 2) (i i )..............................................

34 3.2.5 Natural Phenomenon or Conditibn Threatening Plant Safety (External Threat) 650.72(b)(1)(111) and

$ 5 0. 7 3 ( a ) ( 2 ) ( i i i ).............................................

40 3.2.6 ECCS Discharge into the Reactor Coolant System 9 50. 7 2 ( b) ( 1 ) ( i v)..............................................

44 3.2.7 Loss of Emergency Assessment, Response, or Communications 9 5 0. 7 2 ( b) ( 1 ) ( v)...............................................

46 3.2.8 Internal Threat to Plant Safety

$50. 72(b) (1) (vi) and 6 50. 73 (a) (2) (x)..........................

50 3.3 Four-Hour ENS Noti fications and LER Reports...................

53 3.3.1 Shutdown Plant Found in Degraded or Unanalyzed Condition 9 5 0. 7 2 ( b) ( 2 ) ( i )...............................................

54 3.3.2 Actuation of an Engineered Safety Feature or the Reactor Protection System 650.72(b)(2)(ii) and

$ 5 0. 7 3 ( a ) ( 2 ) ( i v )..............................................

56 3.3.3 Event or Condition That Alone Could Prevent Shutdown of the Reactor, Removal of Residual Heat, Control of the Release of Radioactive Material, or Mitigation of the Consequences of an Accident

$50.72(b)(2)(i11), $50.73(a)(2)(v) and

$ 5 0. 7 3 ( a) ( 2) ( v i )..............................................

65 3.3.4 Common-Cause Failure:; of Independent Trai ns or Channel s $50.73 (a) (2) (vii)..........................

79 3.3.5 Airborne or Liquid Effluent Release 550.72(b)(2)(iv),

9 50. 73 (a) (2) (viii) and 9 50. 73 (a) (2) (i x).......................

83 3.3.6 Contaminated Person Requiring

Transport to Offsite Medical Facility 6 5 0. 7 2 (b) ( 2) (v)...............................................

86 3.3.7 News Release or Other Government Notifications

$ 50. 7 2 (b) ( 2) (vi )............................................. 87 3.3.8 Spent Fuel Storage Cask Notifications

$ 50. 72 (b) ( 2) (v i i ).............................................

92 3.4 Fol l owup No ti fi cati on.........................................

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3.4.1 Followup Reports..............................................

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EMERGENCY NOTIFICATION SYSTEM REP 0RTING.............................

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4.1 Emergency No ti fi cati on Sys tem..............,.................. 96 4.2 Gene ral EN S Reporti ng......................................... 98 4.2.1 Reporting Timeliness..........................................

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4. 2. 2 Reporting Compl e teness........................................

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4. 2.3 Vol untary Noti fi cati ons.......................................

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4. 2.4 ENS Noti fi cati on Retracti on...................................

98 4.3 Typi cal ENS Reporting Is sues..................................

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LICENSEE EVENT REP 0RTS..............................................

101 5.1 LER Reporti ng Gu i del i nes..................................... 101 5.1.1 S u bmi s s i o n o f L E R s........................................... 101 5.1.2 LER Forwarding Letter and Cancellations......................

101 5.1.3 Report Legibility............................................

102 5.1.4 Exemptions...................................................

102 5.1.5 Voluntary LERs...............................................

102 5.1.6 Suppl emental Informati on and Revi sed LERs.................... 103 5.1.7 S p e c i al R e p o r t s.............................................. 10 4 5.1.8 Appendix J Reports (Containment Leak Rate Test Reports............................................

104 5.1.9 10 CFR Part 21 Reports.......................................

104 5.1.10 10 CFR 73.71 Reports.........................................

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5.1.11 Avai l abi l i ty of LER Forms...'................................. 105 5.2 LER Content Requi rements and Preparation Guidance............. 105 5.2.1 Narrative Description or Text (NRCiForm 366 A, Item 17).....................................

107 5.2.2 Abstract (NRC Form 366, Item 16)..............................

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5.2.3 Event Title (NRC Form 366, Item 4)............................

117 5.2.4 Other Fields on the LER Form..................................

118 APPENDICES A Historical Perspective on Event Reporting i

B Emergency Notification System Process C Licensee Event Report Review Programs Second Draft, vii NUREG-1022, Rev. 1 i

D 10 CFR 50.72 Including its Statements of Consideration 4

i E 10 CFR 50.73 Including its Statements of Consideration j

F 1992 Revision to 10 CFR 50.72 and 50.73 Including its Statements of Consideration TABLES 1.

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

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Example Systems.......................................................

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i FOREWORD This second draft of Revision 1 to NUREG-1022 is a result of considerable effort, on the part of NRC staff and public commenters, aimed at developing sound and useful reporting guidance within the scope of the existing reporting rules.

It accommodates many, but not all of the comments that were provided by industry and staff.

4 The principles that underlie the existing rule and revised guidance are:

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Report emergency conditions to State and local authorities and the NRC as quickly as possible to facilitate response and support.

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Report plant-specific safety matters to facilitate NRC followup of corrective actions.

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Report matters that may benefit other utilities, so that they can learn from the experience.

Consideration of these principles led to rejection of an industry comment which opposed guidance for " voluntary reporting."

Based op the comments, certain specific guidance has been deleted. However, because a rule and guidance cannot foresee every circumstance it is important to articulate an industry and regulatory responsibility to report matters that may benefit health, safety, and security.

In doing so, the NRC staff clearly understands the difference between an enforceable legal requirement and a matter of voluntary reporting.

In order to underscore this point, additional guidance will be provided to the NRC staff regarding the non-enforceability of voluntary reports if and when the guidance contained in this Revision 1 becomes final.

The NRC staff provided comments strongly supporting the need for added guidance on reporting human performance aspects of events and conditions.

I Although the statement of considerations for 50.73 specifically addresses reporting of causes and human errors, the suggested guidance went beyond 2

existing requirements. Since a better understanding of the impact of human performance upon risk is the remaining frontier, it is anticipated that improvements in collection and analysis of data related to human performance must occur. However, further development is needed which is outside the scope of this reporting guidance document.

4 Second Draft, ix NUREG-1022, Rev. 1

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 Regulatfons, Part 50, Sections 50.72 and 50.73 (10 CFR 50.73), which became effective on January 1, 1984.

i, Section 50.72 provides for immediate notification requirements via the emergency nc;ification system (ENS) and Section 50.73 provides for 30-day written licensee event reports (LER).

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 j

the industry.

Experience has shown that the threshold of reporting has not-been consistently implemented and some probl. ems exist with the interpretation of the guidelines and definitions. A 1990 survey on the effect of NRC regulation on nuclear.

1 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, Second Draft, xi NUREG-1022, Rev. 1

l importance of reporting timeliness and completeness, and typical NRC concerns associated with ENS notifications for each reporting requirement.

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ACKNOWLEDGMENTS i

l The authors of this second draft wish to acknowledge the members of the task group that prepared the first draft of Revision 1:

J.R. Boardman, P.E. Bobe, l

M.R. Harper, J.L. Crooks, L.M. Padovan, and R.A. Spence. They were assisted 1

by Roger Woodruff of the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Reactor Regulation, Eric Weiss of the NRC Office for Analysis:and'

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' Evaluation of Operational Data, and Geary Mizuno of the NRC Office of the-1

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General Counsel, Dee Gable of the NRC Office lof Administration, and Aimee 1

i Brown of the NRC Office for Analysis and Evaluation of Operational Data; With-regard to this second draft, the authors also wish to acknowledge the legalf advice of Maria Schwartz and Geary Mizuno of the-NRC Office of General j

Counsel, the editing assistance of Dee Gable of the-NRC Office'of Administration, and the word. processing assistance of Shirley Rohrer of the i

NRC's Office of Analysis and Evaluation of Operational Data.

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Second Draft, 4

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I ABBREVIATIONS AE0D 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 cf Federal Regulations CRDM control rod drive mechanism CRVS control room ventilation system DBDR design-basis documentation review DDR design document reconstitution EAR Events Assessment Branch ECCS emergency core cooling system EDG emergency diesel generator EIIS Energy Industry Identification System ENS emergency notification system E0 emergency officer i

E0F emergency operations facility E0P emergency operating procedure EPA Environmental Protection Agency (U.S.)

ESF engineered safety feature (s) l ESW emergency service water FEMA Federal Emergency Management Agency FFD fitness for duty FSAR final safety analysis report H00 headquarters operations officer HP health physics HPCI high-pressure coolant injection HPI high-pressure injection HPN health physics network IEEE Institute of Electrical and Electronics Engineers IIT incident investigation team ILRT integrated leak rate test IN information notice INP0 Institute df Nuclear Power Operations Second Draft, xv NUREG-1022, Rev. 1

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i ISI inserviceinspectidn IST inservice testing LCO limiting condition for operation LER licensee event report, MPC maximum permissible concentration MSIV main steam isolation valve 4

NPRDS nuclear plant reliability data system NRC Nuclear Regulatory Commission (U.S.)

l NRR Nuclear Reactor Regulation, Office of 1

OCR optical character reader a

PDR Public Document Room j

PGA policies, guidance, and administrative controls RBVS react'or building ventilation system.-

RCS reactor coolant system RD0 regional duty officer RHR residual heat removal ROAB Reactor Operations Analysis Branch l

RPS reactor protection system SALP systematic assessment of licensee performance SAR safety analysis report 4

S/D shutdown SIS safety injection system S0V solenoid-operated valve SPDS safety parameter display system i

SR0 senior reactor operator j

STS standard technical specifications TS technical specification (s)

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1 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 2

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 i

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 Backaround 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 i

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

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 j

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 4

events to the NRC Operations Center within a timeframe established by the relative importance of the events.

Events are categorized as either emergencies (immediate notifications, but no later than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ) or i

nonemergencies.

The latter is further categorized into 1-hour and 4-hour notifications; events requiring 4-hour notifications generally have slightly i

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.

Reportinh of nonemergency 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, Second Draft, 1

NUREG-1022, Rev. 1

its root causes, safety assessments, and corrective actions are available, to permit NRC engineering analyses and studies.

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 4

interpretations and directions on certain subjects have been issued in NRC l

bulletins, information notices, and generic letters.

1 1.2 Reportino 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 4

j operational data.

The operational experience submitted in accordance with 10 CFR 50.72 and 50.73 2

is publicly available and has been used by other organizations in ways that 4

are most often beneficial to nuclear safety.

However, uses in areas that were unintended, such as in prudency and reasonableness 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.ich uses, there has been a tendency to only count the number of reported events without i

assessing their individual safety significance.

Such misuses could result in i

licensees adopting a more restrictive reporting threshold in order to reduce I

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

This can be counterproductive to the purpose 4

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 been incurred 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 regu,lation on nuclear power plant activities and subsequent event reporting workshops, identified the need for further guidelines on the two reporting-rules.

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NUREG-1022, Rev. 1

1.3 Revised Reportino 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 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 reportability.

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 Statements of Consideration 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-1022, 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 synopsized 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 i

and voluntary reporting, legal reporting requirements, and human performance issues.

Section 3 contains guidelines on event reportability on specific criteria in 1

i 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.

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NUREG-1022, Rev. 1 W

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.

I 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 950.54(x), and TS prc5ibited operations or conditions are reiterated.

Plant operation in a degraded or unanalyzed condition, or outside the plant's operating and 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 4

discussed. The discussion of LER reporting of common-mode failures of independent safety system trains defines a number of terms and notes their 4

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.

Second Draft, 4

NUREG-1022, Rev. 1

e 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, i

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.

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 i

uses of LERs nationally and internationally.

Appendix D contains 10 CFR 50.72 including its Statements of Consideration as published in the Federa? Register.

Appendix E contains 10 CFR 50.73 including its Statements of Consideration as published in the Federal Register.

Appendix F contains 1992 revisions to 10 CFR 50.72 and 10 CFR 50.73 ir.cluding i

the Statements of Consideration 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.

Reportability Determination l

e 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 requirement;, it 4

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 i

each paragraph. The differences between the ENS and LER reporting requirements are underlined. The differences are discussed when they J

are important.

Key terms are defined and important concepts are Second Draft, 5

NUREG-1022, Rev. I 1

4 1

identified in the " Discussion" sections.

Ev nts used as examples may be-reportable under other criteria but are usually only evaluated for reportability under the specific criter.ia 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, 20.403, 20.405, 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.

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 950.72 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 e

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 tie found in Section 5.2.

f!ew or different guidance e

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.

l l

l i

Second Draft, 6

NUREG-1022, Rev. 1

febte 1 Cceparability of 10 CFR 50.72 and 50.73 Criterle*

Emergency Notification System (ens) (10 CFR 50.T2)

Event or Condition NUREG-1022, Rev. 1 Notification es soon as practical Notification as soon as prectical 30-Day LER Report (10 CFR 50.73) section(s) and in ett cases, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and in att cases, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months EMERGENCY CLASS termediately af ter notification of State Note-Atthough not specificetty 3.1.1 and tecet authorities, but no teter mentioned in -10 CFR 50.73, many then 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months efter dectoration of emergency etess evente involve emergency class defined in Licensee's reportebte ettuations.

emergency plan (50.72(a)(1),(e)(2),(a)(3) and (e)(4)).

TECHNICAL $PECIFICAft0NS (TS):

Plant shutdown (5/D) required Initiation of S/D required by TS Completion of 5/D regJired by TS 3.2.1 a

by TS 150.72(b)(1)(1)(A))

(50.73(a)(2)(1)(A))

13 prohlbited operations or Operation or condition prohibited 3.2.2 condition by is (50.73(a)(2)(1)(3))

is devletion authortred by Devletim from TS authertred by Criterion (50.73(a)(2)(1)(C))

3.2.3 50.54(x) 50.54(x) 150.72(b)(1)(l)ts))

some as EMS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months DECRADED/UNANALYZED CONDtil0NS/0UTSIDE DESIC#

SA515/NOT COVERED BY OPERAi!NG AND ESERGENCY PROCEDURES:

Plant, including its principet During operation, serious degradation Event foi.rld while reactor was W ile either in operation or $/0, 3.2.4, 3.3.1 safety barriers, seriously of plant including its principet safety shut down; had it occurred in condition of plant, incitsfing its 2

degraded barriers (50.72(b)(1)(II))

operation, would have resulted in principal safety barriers, the plant, including its seriously degraded rn principet safety berriers, belts (50.73(e)(2)(ll))

CD seriously degraded O[

(50.T2(b)(2)(l31 on NO N IS

CL MC (D 9

< DJ

=

w.e+

~.

- =. -. -. ~.,

I l

Table 1 (continued)*

'm (ENS) (10 CFR 50.72]

NUREG-1022, l

Emergency Notificatico t Event or Condition Rev. 1 301ey LER Report U0 CFR 50.731 Section(s)

Notification as soon en practicet Notification as soon as practicet ord in at t cases, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and in att cases, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months I

i Plant in ananalyzed condition During operation, plant in unenetyred Event fomd mAlte reactor uns White either in operation or 3.2.4, 3.3.1 significantly compromising condition, significantly c mpromising shut doun; had it occurred in S/D, plant in amenetyred plant safety plant saf ety (50.72(b)(1)(II)( A)]

operetten, would have resulted in condition, significantly the plant being in en amenetyred compromistre plant safety l

condition that significantly (50.73(a)(2)(ll)(A))

l compromises plant selety (50.T2(b)(2)(l)1 t

Plan: stside design bests of During operetten, plant in condition White either in operetton or 3.2.4, 3.3.1 l

plant outside design bests S/0, plant mes in conditten

[50.72(b)(1)(ll)(e)3 (50.73(a)(2)(li)(831 Plant in emittlen not covered During operation, plant in cordition White either in operation or 3.2.4, 3.3.1 by operating and emergency not covered by operating and S/D, plant was in condition not precedares emergency procedares covered by operstlns and

[50.72(b)(1)(li)(C)]

emergency procedJres (50.73(a)(2)(ll)(C)]

CD EX1ERNAL TdREAT TO PLANT SAFETY Any naturet phenomenon or other externet Criterion (50.73(a)(2)(lli)1 3.2.5 condition that poses an actual threat to same es EWS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months l

the safety of the plant or significantly hospers site persomet in performance of dJtles necessary for its safe operation (50.T2(b)(1)(lit)1 ENERGENCY CORE COOLING SYSTEM A vetid ECCS signet that results, or Manuel or automatic actuotton of Criterion (SS.73(a)(2)(lv))

3.2.6, 3.3.2 (ECCS) OtSCHARGE; ACTUATION OF should have resulted, In ECCS any ESF, including the RPS, encogesses both ENS 1 hour and ANT ENGINEERED SAFETY (EST),

discharge into the reactor cootent occurs and was not preptemed as 4 houre INCLLJDlhG REACTOR PROTECTION system (50.72(b)(1)(lvil part of a test or reactor l

SYSTEM (RPS) operation (50.72(b)(2)(III)

EVENTS TRAT ALONE COULD NAVE Event er condition elone would Criterion (50.73(a)(2)(v)] same 3.3.3 l

z PREVENTED FULFILLMENT OF A have prevented futfitteent of as EMS 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Events may C

SAFETY FUNCit0N safety ftnction of system needed include procedural errors, I

for S/D of the reactor, equipment feltures o; discovery a

maintenance of a safe S/D of design, enetysis, e f.n condition, residust heat removal fabrication, construction,

~'@

(RM), control of reteese of and/or procedaret inadequecles.

radioactive material, or Need not report IndividJet NO !

N3 mitigetton of the consequences of cogonent feltures inder this en accident (50.72(b)(2)(Ill))

peregraph if redadent espalpeant

x3 o in same system was operable end g1 evellebte (50.73(a)(2)(vi)1 r+

e-* *

~-

.. ~.. ~.-.....-

~_

-......-~.- -

- -~.---~.-

Tabte 1 (contiroed)*

Emergency Notification Systaan (ENS) (10 CFR 50.721 Event or Cordition NUREG-1022, Rev. 1 Notification as soon as practicet Notification es soon es practicet 30-Day LER Report (10 CFR 50.731 Section(s) and in at t cases, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ord in att cases, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ColoquN CAUSE OR CONDIT!DN RESULTING IN IkDEPENDENT TRAINS Single cause or condition caused 3.3.4 Inoperability of at least one OR CHANNELS SECOMING INOPERABLE Independent train or channet in auttlple systems or two independent trains and channets in a single system designed for safe S/D, RNR, redletion release control, or accident siilgation (50.73(a)(2)(vil)I RADIDACTIVE RELEASES:

Airborne radioactivity reteeses Airborne radioactlvity reteesed Criterion (50.73(e 6 )(vill)(A)2 3.3.5 to en enrestricted eree exceeds seee as ENS 4 houq.

2x the timit specified in 10 CFR 20.1-20.601 Appendia e, Tabte II, to or 20s the concentration specified in 10 CFR 20.1001-20.2401, Appendia e, febte 2, everased over 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (50.72(b)(2)(lv)(A)).

Lipid offluent reteneed to en criterion (50.73(a)(2)(vill)(a))

3.3.5 unrestricted ores eaceede 2x the s:som es Eus 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

limiting caobined maalena permissionable concentration in 10 CFR 20.1-20.6014 Appendix E, Tabte II or 20m the concentration specified En 10 CFR 20.1001-20.2401, Apperdis B, Table 2, for

{

att radionuctides escept tritisse -

end dissotwed noble gases, y

everaged over 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months rn in (50.72(b)(2)(lv)(s)).-

b$

INTERNAL TNREAT TO PLANT SAFETY Any event that posee en actunt -

on Criterton (50.73(a)(2)(x)1 ease -

3.2.8 threat to the safety of the plant

. y@

er algnificantly hospers alte as ENS 1 hour C:L persormet In the conthet of safe g

operetten ISO.72(b)(1)(vi))

.to 1

< Ih7 r+-

. we

~n

~

.~

n_..

l Tebte 1 (continued)*

Emergency Notification System (ENS) (10 CFR 50.721 NUREG-1022, Event or condition Rev. 1 304ey LER Report DO CFR 50.733 Sect W s)

Notification as soon as practical Notification as soon es practicet and in et t cases, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ord in att cases, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LOSS OF EMERGENCY ASSESSMENT, A enjor loss ef capability occurs 3.2.7 OFFSITE RESPONSE, OR for emergency essessment, offsite ColetJNICAT10NS CAPAgILITY response, or communications (50.72(b)(1)(v)]

TRANSPORT OF CONTAMINATED A redloectively contamined 3.3.6 PERSON TO OFFSITE MEDICAL person is transported to en FACILITY offsite medical facility (50.72(b)(2)(v))

NEWS RELEASE /0TNER GOVERNMENT A news reteese la planned or 3.3.7 NOTIFICATIONS other government egencies have been or ulti be notifled of en event rotated to the heetth and safety of the gablic or onsite personnet, er the protection of y

the envirorusent (50.72(b)(2)(vi)1 DEGRADED SPENT FUEL STORAGE A defect In any spent fust 3.3.8 CASK OR CONFluCMENT SYSTEM storage cask structure, system, or component that is leportent to safety (50.72(b)(2)(vil)(All. A significent reduction in the effectiveness of any opent fuel storage cask confinement system during use of the storese cask mder a generet license issued under 10 CFR 72.210

[50.72(b)(2)(vil)(g))

l f

This table is not intended to be used for reportability decisions. ~ Its purpose is to litustrate the compteenntary nature of many IndividJet 7

10 CFR 50.72 and 50.73 criterle and their applicable references in thle report.

i c

l A

Note: FOLL0 LAP MOTIFICATION (SECTION 3.4).

i in Af ter making a 1-hour or 4-hour notification, ticensees are regaired to lunediately notify the NRC operations Center if any of the foltoulne occurs w to OO plant conditions worsen (50.72(c)(1)(1)1, emergency etessification changed 't$0.72(c)(1)(ll)), or emergency etess terminated (50.72(c)(1)(lit)1; ro ro n the results of ensuing evolustions or essessments of plant condittens are obtelned (50.72(c)(2)(l)3; Q-the ef fectiveness of response or protective enesures taken becomee knom (50.72(c)(2)(lill; Infonestlen related to plant behowtor is net ederstood (50.72(c)(2)(lii));

i x (-)

l to -s 7 %-

In addition, if recpaested by the NRC, mainteln en open, continous cosamicetion chercet with the mRC operations Center (50.72(c)(3)).

w.r+ -

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 Enaineerina Judament 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.

1 a

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. Th itsf f@thap(Mth;iMlfi[dMi 6QiWyM{E@jM6jjsentjdj@@@

n{dj jpijMjM6uj@

process 1 supports @heijudgment3j 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 hot 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 Reportina 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).

Second Draft, 11 NUREG-1022, Rev. 1

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. Report all 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.

i l

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 Durina Desian-Bases Documentation Reviews, Safety System Functional Inspections and Other Licenste Encineerina Reviews As indicated in NUREG-1397, "An Assessment of Design Control Practices and 4

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 i,

i reconstitution (DDR) programs, design-bases 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 handle reporting 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 Enaineered Safety Features Actuations There is no standard definition of what constitutes an engineered safety feature. The reporting criterion was based on each plant having defined systems as ESF (e.g., in the plant's final safety analysis report (FSAR)). :16

.ordeW63folh6telcon si j{Ent2efoEtW f6EMnij nimU.niisst(6ff sys tsm5MthsisTsf f requeststthatRicenseesereportgonM voltintary'!basisliffneedi be,Wactuationse of the!Wstjnsfl {' stedd niTjkifej 26 Se eti on.E313 ; 2 y $ss"Ss6tl on '3!3: 2^ foFfu FtheF ^

discussion of this matter.

2.6 Events and Conditions Initially Discussed with the NRC Staff or Identified by NRC Inspections Some licensees personnel have erroneously believed that if a reportable event or condition had been discussed with the resident inspector or other NRC Second Draft, 12 NUREG-1022, Rev. 1

staff, there was no need to report under 10 CFR 50.72 and 50.73 because the I

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.

2.7 Multiple Component Failures There have been cases in which licensees have not reported multiple, sequentially discovered failures of systems or componehts 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 ss50.72(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 l

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 failurs; 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.

A single component failure in a safety system is reportable if it is determined that the failure mechanism could reasonably be expected to occur in one or more redundant components and thereby prevent fulfillment of the system's safety function.

In addition, 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.

Relief Valve Testina When performing periodic surveillance tests of safety or relief valves it is not uncommon to find more than one valve to be lifting outside of the TS-allowed tolerance band, which is typically plus or minus 1 percent.

Rhi6tINifi6Flib)sjahifsE6)l50D2 (b)T2f(TO )?sndf 50lmiR2;}]VH[iysEtsiff c4Hdi t ionsth a tfal onM coul.d fpreven tiful fi l l mentM fga#s'a fet s i tnit i onwo ul d d til l i_s su al lj t be r rep ^o rta 61 eiu nde rc s 50?73(y;fu n a) 2) n"inultijilE~'

{Aunyfaildf5)fbedissi~th'e'existe6Ee df~illriiiTiFdisesiiin'cie('s ~i(Vii)R{commori Second Draft, 13 NUREG-1022, Rev. 1

independent valves is a good indication that the discrepancies probably arose from a common cause. This common cause failure criterion is discussed in Section 3.3.4 of this report.

1 An example involved the sequential testing of-main steam safety valves. Of.

the 20 valves tested,17 were out of tolerance (13 with set points above the technical specification by as much as 4 percent). The licensee initially did-not report this condition because it believed the valves could fulfill their safety function because no safety relief valve set pressure exceeded 1397 psia (110 percent of the system design pressure).

However, the licensee determined a common-mode failure mechanism was the cause for most of the failures; therefore, the condition was reportable as a common mode failure.

~ NigiyBYEiSTEN6H'e7E9"23fi)"(!T(@2 of}R(fii)s FsiiBFf,6 RitT6 HiTsTsffsifiin?il JT!j Ai""diiEuisid~G $iEflon 3.2.

hMity

$pdijip@iesYound,i discrepanc T5 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, to believe that the discrepancy occurred earlier.

However, in the cases of interest here, the existence of similar discrepancies in multiple valves is a good indication that the discrepancies arose over a period of time.

Depending on the significance of the discrepancies and the exercise of engineering judgment, this cituation also may be reportable under one or more of the following sections:

1.,

Section 50.73(a)(2)(ii), seriously degraded, unanalyzed condition that significantly compromised plant safety, outside design basis or in a condition not covered by procedures. These four criteria are discussed in Section 3.2.4 of this report.

2.

If discovered during operation, Section 50.72(b)(1)(ii).

These are the same four criteria as above, discussed in the same section of this report.

3.

If discovered when shut down, Section 50.72(b)(2)(i), seriously degraded or unanalyzed condition that seriously compromises plant safety. This involves only two of the four criteria discussed above. This reporting requirement is discussed in Section 3.3.1 of this report.

Frequently, during an outage, safety valves are removed and replaced with refurbished valves. Then the surveillance testing, on the valves that were INMMd, Mthentisty[esult@indiditedhan%

Tils uldETJiiis IEdiIM5IENN dIIEEfMIE5Ikfdk8is performed later in a shop or test f aci idh removed A$$

ponditj%16]is;sughjonithht@phlhyggt@fslfillmggigaj$

tli exampieg la l,Qj pppd Second Draft, 14 NUREG-1022, Rev. 1 u

J

.A 3

2.8 Human Performance Issues 1

Human performance often influences the outcome of nuclear power plant events.

Detrimental personnel errors may be caused by inadequate procedures, training, l

verbal communications, human engineering, quality control management, or i

supervision. A specific description of the causes and effects of human performance as they relate to an event are to be included in the LER pursuant i

to s50.73(b)(2).

See Section 5.2.1(2) of this report for further discussion of this matter.

i 2.9 Voluntary Reportina

=

The Statement of C'onsiderations for 10 CFR 50.73 specifically addresses the j

use of voluntary LERs.'

It is stated that "... licensees are permitted and encouraged to report any event or condition that does not meet the criteria contained in 950.73(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 1

potential safety significance." The Commission encourages voluntary LERs i

rather than information letters or 10 CFR 50.9 oral reports 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 i

support NRC review of the event and facilitates administrative processing, including data entry. The NRC recognizes that the number of LERs is not in 2

j itself an accurate or appropriate measure to judge a plant's safety i

performance.

Voluntary reporting of LERs is further discussed in Section 5.1.5 of this report.

In addition, voluntary reporting is encouraged under 10 CFR 50.72, as discussed in Section 4.2.3 of this report.

l 2.10 Retraction / Cancellation of Event Reports l

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 1icenYeh so~ chooses,"an ENS'not'iff ati'oh can b'e7et' Fact'edTnd 'an' LER~cih be canceled using the same procedure by which the initial report was made.o'~'

i The retractions' and cancellations are further discussed in Section '4 for ENS i

notifications and Section 5 for LERs. Sound,'" logical' bases for'the withdrawal' l

should be communicated with the request.

(Example 3 in Section^ 3.3'1 '

i illustrates a' case where there"were ~sotind reasons for a retraction. The last i

event under Example 1 in Section 3.3.2 illustrates a case where the reasons l

for retraction were not adequate.)

)

'48 FR 33853, July 26,1983.

i Second Draft, 15 NUREG-1022, Rev. 1 I

a

l 2.11 Time limits for Reportina 10 CFR 50.72 Reporting times in 10 CFR 50.72 are keyed to the occurrence of the event or condition.

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:...."

10 CFR 50.73 10 CFR 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, as was discussed in the guidance in NUREG-1022, SJoplement 1, Question 14.5, if a technician sees a problem, but a delay occurt 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 nroi>1em. Thus, for a single event or condition, it is possible to have several applicable dates:

1.

The Event Date when the event actually occurred (entered in Item 5 of the LER) 2.

The Discovery Date when someone in the plant recognizes that the event has occurred (starts the 30-day clock and should be entered in Item 5 cf the LER (event date) if the event date cannot be clearly defined).

3.

The Report Date when the LER is submitted (entered in Item 7 of the LER).

The previous guidance in NUREG-1022, Supplement 1, Question 14.5, also discussed a "reportability" date, i.e., the date when someone decides or

" discovers" that the event is reportable; however, this date is not used on the LER form or for starting the reportability clock.

Second Draft, 16 NUREG-1022, Rev. I

l If there is a significant leng'th of time (> 30 days) between the event date and either (1)Ithe discovery date or (2) the date when the event was determined to be reportable, the reason for the delay should be discussed in the LER text.

General 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 r rtable. T

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Second Draft, 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.

4

'l l

Second Draft, 18 NUREG-1022, Rev. I l

j

3.1 10 CFR 50.72 and 50.73 General Reauirements 3.1.1 10 CFR 50.72 Immediate Notification Requirements for Operating Nuclear Reactor 50.72(a) General Requirements' 10 CFR 50.73

.[If the event or

"(1) Each nuclear power reactor licensee licensed condition that was under 950.21(b) or 650.22 of this part shall the basis for the notify the NRC Operations Center via the Emergency Emergency Class Notification System of:

declaration met one.

(i) The declaration of any of the Emergency or more of the. 10 Classesspecifipdinthelicensee's-approved CFR 50.73 reporting-Emergency Plan; or criteria, an LER is (ii) Of those non-Emergency events specified required.]

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 methodwhichwillensurethatareportismadegs soon as practical to the NRC Operations Center.

(3) The licensee shall notify the NRC immediately after notification of the appropriate State or j

local agencies and not later than one hour after i

the time the licensee declares one of the Emergency Classes.

"'Other requirements for immediate natification of the NRC by licensed operating nuciear power reactors are contained elsewhere in this chapter, in particular, s 20.205, 20.403 or, for licensees implementing the provisions of 59 20.1001-20.2401, 20.j906,20.2202,50.36,and73.71.

These Emergency Classes are addressed in ApppndixEofthispart.

Commercial telephone number of the NRC Operations Center is (301) 951-0550."

l Continued on next page.

i Second Draft, 19 NUREG-1022, Rev. 1

50.72(a) continued (4) The licensee shall activate the Emergency Response Data System (ERDS) 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 activitated by the licensee during drills or exercises if the licensee's computer system has capability to transmit the exercise data.

(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."

' Requirements for ERDS are addressed in Appendix E,Section VI.

Discussion Appendix E,Section IV (C), " Activation of Emergency Organization," to 10 CFR Part 50, 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, NUMARC/NESP-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 declaring emergencies in the ENS notification will aid the NRC to recognize the significance of an emergency.

The Commission recognized the importance of notification to the NRC of an emergency and amended its regulations without prior notice and comment on February 28, 1980, to require it. Timeframes specified for notification in 950.72(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.

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Second Draft, 20 NUREG-1022, Rev. 1

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1 l

2Notification of the State and local emergency response organizations should be made in accordance with the arrangements made between the licensee and offsite organizations.

Second Draft, 21 NUREG-1022, Rev. I

l l

3.1.2 10 CFR 50.73 Licensee Event Report System i

10 CFR 50.72 550.73(a)(1)'

[ Bases for ENS "The holder of an operating license for a notifications (e.g.,

nuclear power plant (licensee)l shall submit a regardless of plant.

Licensee Event Report (LER).for-any event of the status), are the same as type _ described in this paragraph within 30 days.

i 10 CFR 50.73 where the after the discovery of the event. Unless two rules are otherwise specified=in this section, the

-l 3

~

complementary.]-

licensee shall report an event regardless of the~

plant mode or power. level, and regardless of the significance of the structure, system, or 3

l component that initiated the event."

Discussion t

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 structi!re,.

y system, or component that initiated the event, unless'otherwise specified.-

{

These considerations also are implicit in 10 CFR 50.72 where the' two rules are complementary.

i i

l i

4 I

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Second Draft,

]

22 NUREG-1022, Rev.'1' t

)

3.2 One-Hour ENS Notifications and 30-Day LER Reports This section addresses 650.72(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 650.72(a), licensees are to notify the j

NRC as soon as practical and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of the occurrence of any of the events specified in GE0.72(b)(1) and to submit an LER, if specified.

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

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.

I I

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1 1

i 1

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I l

Second Draft, 23 NUREG-1022, Rev. 1

3.2.1 Plant S,hutdown Required by Technical Specifications 550.72(b)(1)(1)(A) 950.73(a)(2)(1)(A)

Licensees shall report: "The Licensees shall submit a Licensee initiation of any nuclear plant Event Report on: "The completion of shutdown required by the plant's any nuclear plant shutdown required Technical Specifications."

by the plant's Technical Specifications."

If not reported as an emergency under s50.72(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 This 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 s50.72 reporting purposes, the phrase " initiation of any nuclear plant shutdown" includes the performance of any action to start reducing reactor power to achieve a nuclear plant shutdown required by TS.

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For example, if at 0200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months a plant enters an LC0 action statement that states, " restore the inoperable channel to operable status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least Hot Standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ," the plant must be shut down (i.e., at least in hot standby) by 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months .

An LER is required if the inoperable channel is not returned to operable status by 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months and the plant enters hot standby.

An LER is not required if a failure was or could have been be corrected before i

a plant has completed shutdown (as discussed above) and no other criteria in 50.73 apply. This includes a situation where the plant is shutdown, the problem is fixed, and the plant is restarted before the shutdown was required by TS.

Second Draft, 24 NUREG-1022, Rev. 1

4 I

Examples 1

l (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 LC0 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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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.

i The licensee made an ENS notification because of the initiation of a TS-j 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 i

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 LC0 to be in hot i

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 TS-required plant shutdown.

(3)

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

Previous guidance in NUREG-1022, Supplement 1, posed the following situations:

Question 1.2:

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 q

seven day period, is that situation reportable?

i Second Draft, 25 NUREG-1022, Rev. 1

~

I 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 ev,ent be reported.

Question 1.3:

e

' 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.

t Second Draft, 26 NUREG-1022, Rev..I

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 Licensees shall report:

requirement.

However, for certain "any operation or condition operations or conditions prohibited by a prohibited by the plant's plant's TS, other reporting requirements may Technical Specifications."

apply, such as 50.72(b)(1)(ii) and (b)(2)(iii); 50.36(c)(1) and (2); 20.403 (20.2202); and 20.405 (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)(1)(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; 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 document.

(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.

(2)

Limiting Conditions for Operation Section 50.36(c)(2) outlines LCOs in TS. Certain TS contain LC0 statements that include action statements 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.

Second Draft, 27 NUREG-1022, Rev. 1

An LER is required if the conditions of an LC0 are not met (e.g., by exceeding action statement constraints).

The LC0 allows a plant a specified time intervr.1 (referred to as the allowed outage time) 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).

If a condition existed for a time longer than permitted by the TS, 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.)

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(3)

TS Surveillance Requirements Section 50.36(c)(3) outlines surveillance requirements in TS.

For the purpose of evaluating the reportability of discrepancies found during TS surveillances, an operation or condition prohibited by the TS existed and is reportable if the time of equipment inoperability exceeded the LC0 allowed outage time. It should be assumed that the discrepancy occurred at the time of its discovery unless there is firm evidence,-

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As discussed in Section 2.7 ef this report, multiple failures may be an indication of a condition that has persisted for some time.

Missed surveillances are reportable when the surveillance interval plus allowed surveillance internal extensior) {e.9u STSsection4.02}plus the LC0 statement time is exceeded, in

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l Second Draft, 28 NUREG-1022, Rev. 1

(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.

l (5)

Administrative Requirements, Including Radiological Controls, Required by Section 6 of the STS, or Equivalent Section 6 of the STS, 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.

Failure to meet such administrative requirements is prohibited by the TS. Whether it is reportable as an LER depends upon whether it results in a condition covered by the LER rule.

If the violation of the administrative requirements of TS results in operations prohibited by TS, then its 4

reportable.

i I

For example, operation with less than the required number of people on shift would clearly constitute operation prohibited by the TS, or operation with a procedure that had not been properly approved would constitute operation prohibited by the TS.

However, if the requirement is only administrative and does not affect plant operation, then an LER is not required; for example, a change in the plant's organizational structure that has not been approved as a Technical Specification change.

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The proposed rule would have required reporting when "a TS action 3

statement is not met."

The wording of the final rule requires reporting "Any operation or condition prohibited by the plant's Technical Specifications."

The Statements of Consideration for the final rule indicate that this change was made to accommodate plants that did not have requirements specifically defined as action statements (48 FR 33855, July 26, 1983).

Second Draft, 29 NUREG-1022, Rev. 1

(6)

Entry into STS 3.0.3 STS 3.0.3, or its equivalent, establishes requirements for actions when an LC0 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 LC0 action statements are not met. Thus, entry into STS 3.0.3 for any reason or justification is reportable.

(7)

Missed or Deficient 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 uivalent covers these testing requirements.

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Fire Protection Systems When Required by TS When fire protection systems are covered by TS (e.g., through an LCO),

they are within the scope of the LER rule.

Examoles (1)

LC0 Exceeded A licensee found a standby component with a 7-day LC0 allowed outage time and associated 8-hour shutdown action statement to be inoperable during a 30-day surveillance test. 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 LC0 allowed outage time and the shutdown action statement time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months were exceeded.

Had the inoperability been identified and corrected within the 7-day LC0 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 extension was also exceeded. The surveillance tests were imn.ediately performed. An LER is required because the time interval, including extensions Second Draft, 30 NUREG-1022, Rev. 1 4

E

h i

i i

permitted by TS, exceeded the TS surveillance interval plus the LC0 t

action statement times.

j (3)

Entering STS 3.0.3 i

With essential water chillers (A) and-(B) out of service, the only j

remaining operable chiller (A/B) tripped. This condition caused the:

i plant to enter STS 3.0.3 for I hour until chiller (A) was restored to service ud the temperature was restored to within TS-limits.. An'LER is required for this event because STS 3.0.3 was entered.

(4)

Administrative Requirements, Including Radiological Controls, Required i

by Section 6. of the STS, or Equivalent i

1 If a control room is operated with less than the required. number of 1

people on shift or~ is operated with'a required procedure that has not i

4 been properly approved, these operations would constitute a' condition or.

I event prohibited by the TS, and as such are reportable.

However, if a-j requirement is only administrative: and does not substantially and i

directly affect plant operation, then an LER is'not required.

i I

i If a change in the plant's organizational structure is made that has'n'ot.

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Regarding radiation controls, those events covered by 10 CFR 20.403 a

(20.2202) and 20.405 (20.2203), should be reported under 10 CFR 50.72 l

and 50.73, as appropriate.

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(5)

Missed or Deficient Tests Required by ASME Section XI IST and ISI, and'

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by STS 4.0.5, or Equivalent i

Examples of 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 -

i Second Draft,.

i 31 NUREG-1022, Rev. 1 j

i-r r

Y position required for the performance 'of their safety function, verifying motor-operated valve stroke times'for both-(open and closed) directions, using.the proper test pressures to properly classify and test active valves and to increase test frequency-subsequent to obtainin' test results that were below certain threshold values. Ni

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l (6)

Fire Protection Systems When Required by TS The licensee rou'ted a hose from a temporary laundry facility through an emergency diesel generator air intake duct, a ventilation duct, and a wall, breaching the fire barriers, and the licensee took no acceptable compensatory action within the required time. frame. An LER is required.

Second Draft, 32 NUREG-1022, Rev. 1

J 3.2.3 Technical Specification Deviation per 950.54(x) 950.72(b)(1)(1)(B) 950.73(a)(2)(1)(C)

Licensees shall report:"Any Licensees shall report:"Any deviation from the plant's Technical deviation from the plant's Technical Specifications authorized pursuant Specifications authorized pursuant to 950.54(x) of this part."

to 950.54(x) of this part."

If not reported as an emergency under 950.72(a), licensees are required to report any such deviation to the NRC via the ENS as soon as practical and in all cases within I hour.

Licensees are required to submit an LER within 30 days.

Discussion 10 CFR 50.54(x) generally permits licensees to ta'ke reasonable action in an

)

emergency even though the action departs from the lkense conditions or plant technical specifications if (1) the action is imrediately 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.

Example 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 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months 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 alone).

Second Draft, 33 NUREG-1022, Rev. 1

3.2.4 Operating Plant in a Degraded or Unanalyzed Condition 550.72(b)(1)(ii) 550.73(a)(2)(ii)

Licensees shall report: "Any event Licensees shall report: "Any event or condition durina operation that or condition that resultM in the result 1 in the condition of the condition of the nuclear power nuclear power plant, including its plant, including its principal principal safety barriers, being safety barriers, being seriously seriously degraded; or results in degraded; or that resultg in the the nuclear power plant being:

nuclear power plant being:

(A) In an unanalyzed condition that (A) In an unanalyzed condition that significantly compromises plant significantly compromised plant safety; safety; (B) In a condition that 11 outside (B) In a condition that was outside the design basis of the plant; or the design basis of the plant; or (C) In a condition not covered by (C) In a condition not covered by the plant's operating and emergency the plant's operating and emergency procedures."

procedures."

If not reported as an emergency under s50.72(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 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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 bases, or in a condition not covered by the plant's procedures, as described in the rule.

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 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."

The discussions below provide further guidance on reportability under these criteria.

Second Draft, 34 NVREG-1022, Rev. 1

(1)

The condition of the nuclear power plant, including its principal safety barriers, being ceriously degraded.

As indicated ia the Statements of Considerations,.this paragraph includes material (e.g., metallurgical or chemical) problem; 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:

l (a)

Fuel cladding failures in the reacter, or in the storage pool, that exceed expected values, or tht are unique or widespread,. or that are caused by unexpected factors, and would involve a release' -

of significant quantities of fission products.

4 (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.

Esiii$1iiE'6f:;'i6vint~s^thitithelstiff'issTdTo^irsideF7siio7ti6li~1f:Visct5r

' coolant 9ystem' cracks' or! breaks include cracks,or, breaks' that':re'sult':16 l eakage lJiramp1' s t of. events.that's the 1 staff;would (cons'iderWgoetabl e ;a s e

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a Second Draft, 35 NUREG-1022, Rev. 1

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i (2)

The nuclear power plant being'in an unanalyzid condition.that significantly compromises plant safety.

As indicated in the State'ments 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-r to miror variations in individual parameters, or to probl. ems 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.

Aoy trivial single failure or minor error in performing.-

serveillance 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,sthese events should be reported only if they involve functionally-safety."pomponents or if they significantly compromise' plant related "When applying engineering judgment, and there is a doubt-regarding whsti?er to report or-not, the, Commission's policy is that licensees should make the report."

"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 i

'48 FR 39042, August 29,1983 and'48 FR 33856, July 26,1983.

548 FR 39042, August 29, 1983.

Second Draft, 36

_NUREG-1022, Rev. 1

l 1

l l

l accumulation of voids that could inhibit the ability to' adequately remove _ heat from the reactor core, particularly under natural-

.irculation conditions, would constitute an unanalyzed condition

{

and would be reportable."'

"In addition, voiding in instrument lines that results in an i

erroneous indication causing the operator to misunderstand the j

true condition of thp plant is also an unanalyzed. condition _and should be reported."

a (3)

The nuclear power plant being in a condition that is outside the design j

basis of the plant.

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j 48 FR 39042, August 29, 1983 and 48 FR 33355,: July 26, 1983.

)

48 FR.39042, August 29,1983 and 48' FR 33356, JulyL 26,1983'.

7 810 CFR Part 50, Appendix A, Introduction and Criterion 35, and Appendix K, Item I.D.1, indicate that a minimum design criterion is suitable redundancy 1

j meeting the singlo-failure criterion, i

.NUREG-1022L Rev. 1

'i Second-Draft, j

37 3

i l'

(4)

The nuclear power plant being in a condition not covered by the plant's operating and emergency procedures.

JhiV'cH t'eH oii po'ints^ tB~eisfit's "whife^"th'i';ilsiit 'is ~in'?a'~cbiiditf6n outside the coverage of its operating and emergency' procedures. "A straightforward examp'le of this type of event Was'the ' accident <at Thris

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RileIsland..

Examples 4

(1)

Maintenance Error 1

The plant was operating at 100-percent power in steady-state conditio i.

Train "B" essential service water (ESW) system was declared inoperable, depressurized and drained for maintenance. Maintenance technicians were dispatched to loosen train "B" expansion joint in the pipe chase room.

The train "A" expansion joint, also located-in the pipe chase room, was loosened by mistake as a result of a labelling error and water leaked from the loosened flange joint. The licensee declared train "A"

ESW system inoperable and entered TS 3.0.3 because both trains of ESW were inoperable. Repairs were initiated to replace and retorque train "A" expansion joint flange bolts. Train "A" ESW system was declared operable and TS 3.0.3 exited before commencing a plant shutdown.

The licensee made an ENS notification under 10 CFR 50.72(b)(2)(ii)(A) as an unanalyzed condition that significantly compromised plant safety.

In a subsequent engineering evaluation the licensee determined that leakage from the loose flange-was insignificant and the flange would remain in place during a design-basis earthquake and, thus, the "A" ESW train was operable and the event was not reportable. However, a voluntary LER was submitted within 30 days.

(2)

Unqualified Component The plant was operating at 100-percent power in steady-state condition.

During a review of component classifications, the licensee identified some non-safety-related components which were connected to the drywell (primary containment) safety-related nitrogen supply header. During efforts to upgrade the components to safety-related in accordance with plant procedures, it was determined that certain parts within the non-safety-related components were made of a material that is not suitable for high temperature conditions.

't appeared that failure of these parts during post loss of coolant accident (LOCA) conditions could result in the depressurization of the nitroget, supply header and lead to the inability to provide a 100-day supply o' nitrogen to safety-related automatic depressurization system (ADS) valves, as described in the updated final safety analysis report (UFSAR). The licensee made an ENS notification because of a condition' that placed the plant outside of its design basis. The licensee determined, based on subsequent engineering evaluation, that the maximum Second Draft, 38 NUREG-1022, Rev. 1

leakage rate would be less than the capacity of the drywell nitrogen supply header valves and the 100-day supply of nitrogen was not adversely affected and, thus, the event was not reportable. The ENS notification was retracted.

Second Draft, 39 NUREG-1022, Rev, 1 l

s

d 3.2.5 Natural Phenomenon or Condition Threatening Plant Safety (External Threat)'

650.72(b)(1)(iii) 650.73(a)(2)(iii)

S Licensee shall report: "Any natural Licensee shall report: "Any natural phenomenon or other external phenomenon or other external condition that poses an actual condition that posed an actual threat to the safety of the nuclear threat to the safety of the nuclear power plant or sianificantly hampers power plant or significantly site personnel i' 'he performance of hampered site personnel in the duties necessary the safe performance of duties necessary for operation of the lant."

the safe operation of the nuclear power plant."

If not reported as an emergency under 950.72(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 s50.72(b)(1)(vi) and s50.73(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 j

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 i

Second Draft, 40 NUREG-1022, Rev. 1

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 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.

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acthgQ5atii@MaMogGyprastipaB4RfnE6@aispisnhipsl})@H In most cases, events such as earthquakes, approaching hurricanes or to ado 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 j

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 SieressuchTthit@ithiperso.hoursnnel zwere; si@61ficanti m

hf%t@[nechs s s FfifoM s sfR0pspati on@For$x.f! h a the situation would be reportable Kith ily t

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Second Draft, 41 NUREG-1022, Rev. 1

Examples (1)

Earthquake Seismic alarms were received in the Unit I 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.

The licensee classified this as an Unusual Event in accordance with the emergency plan and notified the NRC via DS per 650.72(a)(1)(i) within 30 minutes of the earthquake. The licenses terminated the event after walkdowns of the plant were satisfactorily completed and made an ENS update call.

No LER was subnitted 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 issue ! by the National Hurricane Center. The hurricane was predicted to raach the site in approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 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 safety-related structures.

Extensive damage occurred on site. The Unusual Event was upgraded to an Alert when the pressurized fire header was lost because of storm-related 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 i

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 l

lost.

Both station emergency diesel generators received a fast start signal and load sequenced as designed.

Five minutes later, offsite Second Draft, 42 NUREG-1022, Rev. 1 j

i m.

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 burn 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 occurrence of natural phenomenon that posed an actual threat and several other reporting criteria as well.

.f Second Draft, 43 NVREG-1022, Rev. 1

3.2.6 ECCS Discharge Into the Reactor Coolant System 850.72(b)(1)(iv) 10 CFR 50.73 Licensees shall report: "Any event

[ECCS discharge is a subset of s

that results or 'should have resulted 550.73(a)(2)(iv), actuation of an h

in Emergency Core Cooling System engineered safety feature (ESF), as (ECCS) discharge into the reactor discussed in Section 3.3.2.

coolant system as a result of a Therefore, an LER is required.]

valid signal."

If not reported as an emergency under s50.72(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 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Discussion i

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, alth6 ugh 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 test. 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 550.72(b)(2)(ii) requires a report within four hours and s50.73(a)(2)(iv) requires submittal of an LER.

The staff considers deliberate manual ECCS initiations or actuations based on the operatcr'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 i

Second Draft, 44 NUREG-1022, Rev. 1 l

l l

1

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 950.72(b)(1)(iv) also requires a 30-day LER under

$50.73(a)(2)(iv) because an ESF was actuated.

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 enndensate and feedwater systems were restored, these pumps were se, *d.

An ENS notification is required under 950.72(b)(1)(iv) because an ECCS i

system injected water into the RCS as a result of a valid ECCS signal.

Although the RPS actuation also is reportable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months under

$50.72(b)(2)(ii), this more limiting criterion applies. An LER is required under s50.73(a)(2)(iv) because an ESF actuation occurred.

(2)

PWR ECCS Injection following Surveillr..e 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 950.72(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 950.73(a)(2)(iv).

(3)

PWR ECCS Injection Caused by Personnel Error While surveillance testing containment isolation valves, a test pushbutton 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 s50.72(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 650.72(b)(2)(ii) and an LER is required under s50.73(a)(2)(iv).

Second Draft, 45 NUREG-1022, Rev. 1

3.2.7 Loss of Emergency Assessment, Response, or Communications

$50.72(b)(1)(v) 10 CFR 50.73 Licensees shall report: "Any event

[No corresponding Part 50.73 that results in a major loss of requirement.]

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 via the ENS of a major loss of their emergency assessment, offsite response, or communications capability as soon as practical and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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 the staff considers to be a major loss of emergency assessment, offsite response, or communications capability include those in which any of the following are not available:

Safety parameter display system (SPDS) e Emergency response facilities (ERFs) including emergency operating facilities (E0F's) and technical support centers (TSC's)

Emergency communications facilities and equipment including the e

emergency notification system (ENS)

Public prompt notification system including sirens e

Plant monitors necessary for accident assessment e

loss of Emeroency 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 Second Draft, 46 NUREG-1022, Rev. I 1

particular equipment, e.g., loss of only the SPDS for a 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.

85WsisEth"s7siliiiillitiil f(j33f?6IsT66iGfidUhdihWE5 iip 5hintI6RffsTH7IEESIIsT4 metsorologicslitbwerp fadiati6nimonitoFMplsntichmphtsr#ypassfishal5nt? "6r2ERF pesibdid fetimshgeneralifiipnbtWebo rtabl e OF6Mthiiit whi(hfisjefys:pareipf call Ad j uppff ths(s;t sff@ould$djiided@odgfftJjssHiH thag[hoursstojbejshortg loss of Offsite Response Capability A major loss of offsite response capability includes those events tnat 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 facilities, or public prompt notification system, including sirens and other alerting systems.

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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 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months would warrant an immediate notification.

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PIpceduresg Second Draft, 47 NUREG-1022, Rev. 1

'i Loss of Communications Capability A r.ajor loss of communications capability may include the loss of ENS and/or cther 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 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 s50.72(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.

IE a~similir~m^aWnsF;' i f; th~e" Nat > sut pTisd teTeish'66s"l ih'e'6F ':mbdeE*Iised~~f6r*tha i

emergency response data system is~ inoperable, the NRC' operations center 'should be inforued so that repairs can be ordered.( Howevert this, does not ge,ne, rate

.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 l

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 for several hours because of e

loss of power as a result of severe weather.

Second Draft, 48 NUREG-1022, Rev. 1

28 of 54 alert sirens reported out of service for an hour as a result of a local ice storm and a return-to-service estimate was U'iKnown.

All offsite emergency sirens were:

- found inoperable during a monthly test.

- taken out of service for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of repair.

- inoperable because control panel power was lost for an unknown period.

- inoperable because the county radio transmitter failed for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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 l

I 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 procedures or practices to compensate for the lost emergency sirens. An LER is not required.

(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.

(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 I 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.

2

-Second Draft, 49 NUREG-1022, Rev. 1

3.2.8 Internal Threat to Plant Safety 950.72(b)(1)(vi) 950.73(a)(2)(x)

Licensees shall report: "Any event Licensees shall report:

"Any event that poses an actual threat to the that posed an actual threat to the safety of the nuclear power plant or safety of the nuclear power plant or significantly hampers site personnel significantly hampered site in the performance of duties personnel in the performance of necessary for the safe operation of duties necessary for the safe the nuclear power plant including operation of the nuclear power plant fires, toxic gas releases, or including fires, toxic gas releases, radioactive releases."

or radioactive releases."

4 If not reported as an emergency under s50.72(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

)l threats, 550.72(b)(1)(iii) and 650.73(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 In addition, the staff considers the following standards appropriate in this regard:

4 The significant hampering criterion is pertinent to "the performance of duties necessar Ohs

%tbisVhldi@y for safe operation of the nuclear power plant."hiWi s}ti

'48 FR 33856, July 26, 1983.

Second Draft, 50 NUREG-1022, Rev. 1

I I

il i

Significant hampering includes hindering or interfering (such as with-j e

J protective clothing or radiation work permits) provided that the interference or delay is suff,@1cient to significantly threaten the safe tis $ths $iisdisiaQ[is ~

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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.

I j

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 j

of the operators to perform necessary duties is significantly hampered.

Fairly common events such as minor spills, small gaseous waste releases, or the disturbance of contaminated particulate matter (e.g., dust) that require i

temporary evacuation of an individual room until the airborne concentrations decrease or until respiratory protection devices are used, are not reportable 2

unless the ability of site personnel to perform necessary safety functions is

{jf@u];d3EJ significantly hampered.

1 i

No LER is required for precautionary evacuations of rooms and _ buildings that l

subsequent evaluation determines were not required.

Even if an evacuation l

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.

i significantly hampered in carrying out their safety responsibilities.

i Fires pose a unique threat in that (1) until the fire has been extinguished.

i the extent ot~ its spread is open ended and (2) at any time the' full extent of l

damage affecting the safe operation of the nuclear power plant may not-be readily apparent.

In most cases, fires result in ENS notification because there is a. declaration i

of an emergency class, which is reportable under 950.72(a)(1)(ii) as discussed j

in Section 3.1.1 of this report, rather than because the fire is considered to i

I 4

Second Draft, j-51 NVREG-1022, Rev. 1 i

1 1

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constitute an actual threat or significant hampering.'

Often, with the passage of time, it is apparent that an actual threat or significant hampering i

did not occur and, thus, no LER is submitted.

In other cases, the event is l

judged to meet one of these criteria and an LER is submitted.

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thlisloestid d6ssMiotVs igni fiesntlys h'ampenitheispefatorstandid6esinstia6d pould(nbtMs$ssonabli[beisxpshtsd [t6;;thisst edipisht re sortablsfundstjitlyi Q cpi teri6ni j Exsinpleigould$nt10desmallipap;eQifesRh asstrysgomtrashicapsgoMcigare tejburnsioQfurniturejorjdpholsteryj Examples Previous guidance in NUREG-1022, Supplement 1, posed the following situations:

Question 9.4:

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 9.5:

e 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.

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 be considered an Unusual Event.

t Second Draft, 52 NUREG-1022, Rev. 1

1 3.3 Four-Hour ENS Notifications and LER Reports This section addresses 550.72(b)(2), "Non-Emergency Events--Four-Hour 4

Reports," and 10 CFR 50.73 written reports associated with these 50.72 notifications.

If not reported as a declaration of emergency class under 550.72(a) or as a non-emergency 1-hour report under 650.72(b)(1), licensees are to notify the NRC as soon as practical and in all cases within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of the occurrence of any of the events required by 950.72(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, I

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 3.3.7, requires ENS notification but no LER.

i I

i i

1 i

i 1

1 i

Second Draft, 53 NUREG-1022, Rev. 1

3.3.1 Shutdown Plant Found in Degraded or Unanalyzed Condition 550.72(b)(2)(i) 10 CFR 50.73 Licensees shall report: "Any event (Events found while the reactor is found while the reactor is shut shutdown that involve degradation of down, that, had it been found while the principal safety barriers or the reactor was in operation. would unanalyzed conditions that have resulted in the nuclear power significantly compromise plant plant, including its principal safety are addressed by safety barriers, being seriously 50.73(a)(2)(ii).

Therefore, an LER degraded or being in an unanalyzed is required.

See Section 3.2.4.]

condition that significantly compromises plant safety."

If not reported under 950.72(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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 With the plant in Mode 6 (refueling), ultrasonic testing revealed a number of failed fuel rods (approximately 233 were identified in 88 of 109 fuel assemblies scheduled for reinsertion) that far exceeded 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.

Second Draft, 54 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.

l 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 j

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 i

this problem.

As in example (1), and 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 event report and LER were not required. Consequently, after discussion this event with the Regional Office, the licensee retracted this event.

Second Draft, 55 NUREG-1022, Rev. 1

3.3.2 Actuation of an Engineered Safety Feature or the Reactor Protection System 950.72(b)(2)(ii) 550.73(a)(2)(iv)

Licensees shall report "any event or Licensees shall report "any event or condition that results in a manual condition that resultd in a manual or automatic actuation of any or automatic actuation of any Engineered Safety Feature (ESF),

Engineered Safety Feature (ESF),

i including the Reactor Protection including the Reactor Protection System (RPS) except when:

System (RPS), except when:

(A)

The actuation results from and (A)

The actuation resultd from is part of the preplanned and was part of the pre-sequence during testing or planned sequence during reactor operation; testing or reactor operation; (B)

The actuation 11 invalid and:

(B)

The actuation was invalid and:

s (1)

Occurs while the system is (1)

Occurrd while the system is properly removed from service; properly removed from service; (2)

Occurs after the safety (2)

Occurrd after the safety function has been already function has been already completed; or completed; or (3)

Involves only the following (3)

Involved only the following specific ESFs or their specific ESFs or their equivalent systems; equivalent systems; (i)

Reactor water clean-up (i)

Reactor water clean-up system; system; 1

(ii) Control room emergency (ii) Control room emergency i

ventilation system; ventilation system; (iii) Reactor building (iii) Reactor building j

ventilation ventilation system; system; (iv)

Fuel building (iv)

Fuel building ventilation system; or ventilation (v)

Auxiliary building system; or ventilation system."

(v)

Auxiliary building ventilation system."

If not reported under s50.72(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 es pragtical and in all cases within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 1

Second Draft, 56 NUREG-1022, Rev. I

l l

l 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.

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 mitigate the consequences of an event (e.g., at the discretion of the licensee as part of a preplanned procedure). '

This indicates an intent to require reporting actuations of features that

^ Generally, the" staff wo'uld mitigate the consequences of significant events.

not'coiside'r this to inc10de' single ^ conifoniint Fctuations because' single components of complex systems, by themselves,' usually do not mitigate thd s

consequences of significant events. However,.in some cases a component wonid be sufficient to mitigate the event (i.e., perform the ESF function) and 'its thestatementthatthereportingrequirementisbasedonthepremisethat'~j actuation would, >therefore, be reportable. This position is consistent witi ESF's,areprovidedtomitigatethe; consequences _of_asignificantevent.f

$iEgibMi]6sid6fiit tTiiMilisfe5{GijuMuss?[ihM{@iiggi]Hyl^sysMiE{$t}liis are; reportable; In 'this regard, the' staff considers act'uati^oii"6f a ~diefel ~generatoF to bi

~

actuation of a train -not actuation of a single component - because a diesel e

generator mitigates the event (performs,the ESF f, unction)< (See Example 3 belowd Th~e'~ stiff 'also con'sid' rs 'deliberat'e sahdil'" action's, lnWich" oriif or more ESF e

components are actuated in response to actual plant conditions, to, be reportable because such actions would usually. mitigate the consequences of a significant event. This position is consistent with'4 the statement 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 decreasing pressurizer level or starting HPCI in response to a loss of feedwater would be reportable. However, shifting alignment' of makeup pumps or closing a containment isolation valve for normal operational purposes would not be reportable.

The Statements of Considerations also indicate 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 otherwige, are not reportable if they do not complete the minimum actuation logic.

"48 FR 33854, July 28,1983, 48 FR 39043 and 48 FR 39044, August 29, 1983.

48 FR 33854, July 28, 1983, 48 FR 39043 and 48 FR 39044, August 29,

'2 1983.

Second Draft, 57 NUREG-1022, Rev. 1

Note,ThoWever, that. if only a' single' ESFAS Mhannel. actuates ::inLresponselto plant < parameters for which there should have beensan1 actuation,a thisiwould amountitoMa;faildre of the: ESF.,It would generallyLbe? reportable!:underithese criteriaL(ESFactuation) as.well ast under 10 CFRL50J2(b)(2)(iii)Landi10:CFR 50'.73(a)(2)(v)L(eventorcondition~alone).

This: position is consisteht1with the statementethat the Commission.is interestedsin. events!where an ESFJwas x

needed'togitigatettheJconsequencesl,-whether'orfnot;the;equipmentiperformed properlyJ 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.

However, 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 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 develcp during the shutdown that require an automatic reactor trip, such a reactor trip should be reported.

The fact that the safety analysis assumes that an ESF will actuate automatically during an event does 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 discretion of the licensee as part of a planned procedure).',(e.g., at the This implies that the procedural step indicates the specific actuation that will be generated and control room personnel are aware of the specific signal generation before its occurrence or indication in the control room.

However, if the system actuates during the planned operation or test in a way that is not part of the planned procedure, such as at the wrong step, that event i.s reportable.

Note thatlif an operator'were to. manually tripf thefreactoE in a'nticipation of n

receiving an automatic trip, this would be reportable'justJas;the~ automatic trip _ wo_uld 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.

These amendments eliminate reporting of invalid ESF actuation of systems which had been properly removed from service or for which the safety function which the ESF is intended to accomplish had already beer accomplished, Valid ESF actuations are those actuatians that result from " valid signals" or t

j from intentional manual initiation, unless it is part of a preplanned test.

l l

13Also 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.

"*48 FR 33854, July 28,1983, 48 FR 39043 and 48 FR 39044, August 29, Second Draft, 58 NUREG-1022, Rev. 1

Valid signals are those signals that are initiated in response to actual plant j

i 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.

1 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 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 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.

Finally, invalid actuations of certain specified systems are not reportable.

These systems are limited to 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 bt ilding ventilation system or equivalent ventilation sy m ms.

Invalid actuations of other ESF systems continue to be reportable.

For BWRs, the actuation of the standby gas treatment system in response to 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 j

event, in crder to provide a complete, accurate and thorough description of the event.

There are no standard definitions of ESF or RPS.

The reporting criterion is based on each licensee having defined systems as ESF or RPS (e.g., in the plant's FSAR, but not necessarily limited to Chapters 4, 6, and 7). Actuation of a system would be reportable if that system is classified as an ESF or as a portion of the RPS; if not, the actuation is not reportable under this criterion.

If idditi65,^Kn' 6Fdef: t67f6ihot'e" con'sistint'YeV6Ft'iHcf fo?T"m'ihihim'Tet!6f

' safety systems, the staff requests that licensees report / actuation of all ths systems identified:in Table 2., As discus' sed < above,' reporting would be" ~ ~

required if the actuated system is one that:the licensee' h'as classified"ai ari ESF or' part of the RPS.

If this is not'the case,' but: the actuated s stem'W In:luded in Tab 4 2,,the reportirig;would' be, voluntary.f~' '~'~ ~ ^'~~ ' ~ "y l

Second Draft, 59 NUREG-1022, Rev. 1

Table 2 contains systems typically reported under the criterion, by at least some licensees.

Systems not identified in this table should not be misconstrued as unimportant or insignificant because of their omission.

Examples (1)

RPS Actuation The licensee was placing the residual heat removal (RHR) e 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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.

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 (550.72) 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 neither the loss of recirculation pumps nor off-normal procedure entry were 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.

Second Draft, 60 NUREG-1022, Rev. I l

(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.

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Emergency Diesel Generator (EDG) Starts The licensee provided an LER describing an event in which the EDG e

autoratically 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 i

ESF actuation (EDG auto-start at this plant) was not identified at the step in the calibration procedure being used.

I The licensee provided an LER describing an event in which, after e

an automatic EDG start, and for unknown reasons, the emergency bus feeder breaker from the EDG did not close when power was lost on l

the hus. An ENS notification and LER are required because the ESF-actuation logic for the EDG start was completed, even though the j

diesel generator did not power the safety buses.

(4)

Preplar.ned Manual Scram j

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 i

(MSIVs), an operator incorrectly closed MSIV "D" when the procedure specified closing MSIV "C".

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Second Draft, i

61 NUREG-1022, Rev. I

I (6)

Control Room Ventilation System (CRVS) Isolation l

l 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.

J 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 a

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 non-ESF 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.

(8)

Manual Actuation of ESF Component in Response to Actual Plant Condition 9

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.

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Second Draft, 62 NUREG-1022, Rev. 1

(9)

Actuation of ESF 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 potentit.1 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 ESFs prior to start of work.

During the event, a loss of continuity did occur and the ESFs 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.

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staff has concluded'that the eventis not~repoMable' because the event was not, listed as' definitely;goingsto' occur.t it wat' itsd as having a high probability'to occur andtwas documented intapp ste ^

j procedures; (2) ' plant' operating staff l clearly recognized'>the po~tential for the event to':occuri and (3) ino other' unexpected ESFflor:other'~""

gi$atiogoccurledwhichasopLrecognized[andJtated_i[Qpiprocjd@id However, if during a planned procedure or test, the ESF actuates in a way that is not part of the planned procedure or the unexpected ESFs occur, the e. vent would be reportable.

i l

Second Draft, 63 NUREG-1022, Rev. 1

]

l

't Table 2 Example Systems Emergency Core Cooling Systems (ECCSs)

For pressurized water reactors PWRs):

reactor coolant system accumulators e

boron injection system i

e high, intermediate, and low-head injection systems, including systems for charging using centrifugal charging pumps, safety 2

injection, and residual (decay) heat removal and their water j

sources q

For boiling water reactors (BWRs):

high-and low-pressure core spray systems and their water sources high-pressure coolant injection system, feedwater coolant i

injection system, residual heat removal system, and their water sources isolation condenser system, reactor core isolation cooling e

system automatic depressurization system j

Anticipated transient without scram (ATWS) Mitigating Systems Containment Systems containment and reactor vessel isolation systems e

containment heat removal and depressurization systems, e

including the containment spray and additive system and the fan cooler system containment air purification and cleanups systems containment combustible gas control systems, including hydrogen e

recombiners, igniters, nitrogen inerting systems, and containment atmospheric dilution systems BWR standby gas treatment systems Heating, Ventilating and Air condition (HVAC) Systems for the Control Room and Fuel Handling areas PWR Auxiliary Feedwater Systems Electrical Systems emergency ac electrical power systems, including emergency diesel generators (EDGs) and their associated support systems (even if classified as an essential auxiliary support in the plant's safety analysis report Division 3 EDGs and their assoc (SAR), and BWR dedicated iated support systems ac_tuation and control systems (including associated interlocks) for engineered safety feature (ESF) systems Second Draft, 64 NUREG-1022, Rev. 1

i i

3.3.3 Event or Condition That Alone Could Prevent Shutdown of the Reactor, Removal of Residual Heat, Control of the Release of Radioactive j

Material, or Mitigation of the Consequences of an Accident

=

$50.72(b)(2)(iii) 950.73(a)(2)(v)

Licensees shall report: "Any event Licensees shall report: "Any event or condition that alone could have or condition that alone cotid have prevented the fulfillment of the prevented the fulfillment cf the safety function of structures or safety function of structures or i

systems that are needed to:

i (A) Shut down the reactor and (A) Shut down the reactor and maintain it in a safe shutdown maintain it in a safe shutdown l

condition; condition; (B)

Remove residual heat; (B)

Remove residual heat; (C) Control the release of (C) Control the release of 4

radioactive material; or radioactive material; or t

(D) Mitigate the consequences of an (D) Mit* gate the consequences of an i

accident."

i 10 CFR 50.72 953.73(a)(2)(vi)

.i

[The Statements of Consideration for

" Events covered in paragraph 10 CFR 50.72 contain wording similar (a)(2)(v) of this section may to those of 650.73(a)(2)(vi).]

include one or more personnel errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or i

i procedural inadequacies. However,

)

individual component failures need not be reported pursuant to this i

i paragraph if redundant equipment in 1.

the same system was operable and available to perform the required safety function".

If not reported under 650.72(a) or (b)(1), licensees shall notify the NRC via the ENS as soon as practical and in all cases within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of discovery of the event or condition and submit an LER within 30 days.

1 Discussion The level of jiidgment for' repsrting an event or' condition,iindsr thif eriterion

~

is a reasonable expectation of preventing' fulfillment of a safety function; In the discussions which follow, many of which are taken from the' Statement'of

~

Considerations or from previous NUREG guidance, several different expressions such 'as, "would have," "could have," "alone,c_ould. ave,", and ". reasonable doubt" h

Second Draft, 65 NUREG-1022, Rev. 1

'a~rs'used 't6 ^ch^aV acterize' th~i s~'standa'rd.^' In^ th'e" staff's" view,~ all fof thiss should be judged on the basis of'a reasonablgpectation of p,reventing fulfillment of the safety function.

As indicated in the Statement of Cons'.derations, the intent of these criteria is to capture those events where there would have been a failure of a safety werediscoveredorwhetherthesystemwasneededatthetime.'p-system to properly complete a stfuy function, regardless of w en the failures These criteria cover an event or condition where FFdEEd3% structures, components, or trains of a safety system could hail 6~fii1Fd to perform their intended function because of:

ersonnel errors, including one or more p#idiyditifi6Af6fsdisEij or design procedure violations; equipment failures;stib(Ap3]EEEEf"TE f

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to perform the safety function (e.g., high pressure core cooling failed, but feed-and-bleed or low pressure core cooling were available-to prov;de 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."

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$9stemMi t$1 sinotheces(a$typsj0ssjahijddj tOhaQsridsspi tjjjg{ fsijiirRijd thatisysten The term " safety functior." refers to any of the four functions (A through D) listed in these reporting criterit that are required during any plant mode or accident situation as described or relied on in the plant safety analysis raport 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.

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Second Draft, 66 NUREG-1022, Rev. 1

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As indicated in the Statement of Considerations:

"The Commission recognizes that the application of this and other paragt'phs of this section involves the use of engineering judgment.

In this case, a technical judgment most 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 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 j

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 moretrainsshouldbereportedif,inthejudgmentofthelicgsee,the functional capability of the overall system was jeopardized."

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 actvily 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 "48 FR 33854 and 48 FR 33858, July 26, 1983.

Second Draft, 67 NUREG-1022, Rev. 1

1 Train "A" and instead of shutting the pump discharge valve ig Train "A,"

he mistakenly shuts the pump discharge valve in Train "B")."

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.

Single Failure Thest: 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 commor.-mode failure implications). However, a sing'.e 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 system function, has been found acceptable.

As discussed in the St.tcments of Consideration, "there are~a limited number of single-train systems that perform safety functions, such as the BWR high-pressure cuolant injection and reactor core isolation cooling systems that may t,9 taken credit for in the plant's safety analysis report or covered in the technical specifications.

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 pecifications may allow such a condition to exist I

for a limited time."

Common-Cause Failures The following conditions are reportable under these criteria:

an event or condition that disabled multiple trains of a system 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 reasonable doubt that the second train would complete its safety function if called upon an observed or identified event or condition that alone could have prevented fulfillment of the safety function l

1748 FR 33854 and 48 FR 33858, July 26, 1983, 1848 FR 33855, July 26, 1983.

Second Draft, 68 NUREG-1022, Rev. 1

i 1-k I

1 Multiple equipment inoperability or unavailability i

4 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 inoperability or unavailability, it is reportable i

under these criteria. This would include cases where one train is j

disabled and a second train fails a surveillance test.

i 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 i

mode.

It also is independent of whether:

the system or structure was demanded at the time of discovery i

l the system or structure was required to be operable at the time of i

discovery 4

j the cause of a potential failure of the system was corrected before an e

i actual demand for the safety function could occur l

other systems or structures were available.that could have or did.

j perform the safety function the entire system or structure is specified as ESF or safety related, if' 1

the plant safety analysis report relied on it to perform or if it i

supports or could affect a system that performs a safety function the problem occurs in a non-safety portion of a cystem, if it prevents j

the performance of the safety function The following types of events or conditions generally are not reportable under

.l l

these criteria:

i i

failures that affect inputs or services to' systems that have no safety e

function (unless it could prevent the performance of'a safety function of an adjacent or interfacing system)

J 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 i

performing its function) independent failure of a single component (unless it is indicative of a j

e generic problem, it alone could have caused a safety system f ailure, or it is in a single-train system) a procedure error discovered before procedure apm val and the error could have resulted in defeating the system function Second Draft, i

69 NUREG-1022, Rev. 1 i

'j

a failure of a system used only to warn the operator where no credit is e

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 e

fulfillment of a reactor shutdown unrelated component failures in several different safety systems e

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 p'imilar events may, however, be reportable under other sections of the rules.

Examples (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 LC0 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 returned 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 j'

the following situation:

i Previous guidance in NUREG-1022, Supplement 1, Question 7.14, discussed l

At our plant, RCIC is not a " safety system" in that we assume no credit for its operation in our safety analysis. Are failures and unavailability of this system reportable?

i 48 FR 33854, July 26, 1983.

Second Draft, 70 NUREG-1022, Rev. 1 l

1

Answer:

If RCIC is not considered to be an ESF, then its actuation is not reportable under 50.73(a)(2)(iv). However, 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 50.73(a)(2)(v).

If the RCIC is covered under a Technical Specification surveillance test requirement, then an LER is required under 50.73(a)(2)(1)(B) if the Technical Specification is violated.

(3)

Failure of a Single-Train Environmental System Previous Guidance in NUREG-1022, Supplement 1, Question 7.13, discussed the following situation:

1 1

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 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 the system does not perform a " safety function" in the context of the LER rule and failures of the system are not reportable.

(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 (95 *F and 0 psig; before initial criticality) and a post-modification 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 Second Draft, 7I NUREG-1022, Rev. 1

detector voltage was disabled. The input error inhibit switch was immediately returned to the normal position and a caution was added to approp*iate plant instructions.

This event is reportable because disabling the source range detectors could have prevented fulfillment of the safety function to shutdown 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 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 sunnator 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 F1 coding in Saltwater Pump Bay During maintenance activities on the south saltwater pump, the licensee was removing the pump internals 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 Previous guidance in NUREG-1022, Supplement 1, Question 7.1, discussed the following situation:

Some clarification is needed for events or conditions that alone "N A 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 Second Draft, 72 NUREG-1022, Rev. 1

4 the use of the wrong lubricant, the system "could have" or "would have" failed.

1 (9)

Oversized Breaker Wiring Lugs Previous guidance in NUREG-1022, Example C-14, discussed the following 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 Nol 20 AWG, A No.18-22 lug is the acceptable industry standard for a No. 20 AWG wire.

Since the trip coils were supplied pre-wired, all safety-related breakers utilizing the trip coil were inspected. All other breakers inspected had 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 Nol 18-22 AWG lugs.

Comment:

The event is reportable because the incorrpatible pigtails and lugs could have caused one or more safatY systems to fail to perform their intended function [50.73(a)(2)(v)].

(10) Contaminated Hydraulic Fluid Degrades MSIV Operation Previous guidance in NUREG-1022, Ex uple C-48, discussed the following situation:

During a routine shutdown, the operator noted that the #11 MSIV closing time appeared to be excessive. A subsequer.t ' 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 4

stick and delay the transmission of hydraulic pressure to the actuator.

Three more filters will be purchased providing supplemental filtering d

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)].

Second Draft, 73 NUREG-1022, Rev. 1

4 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 The previous guidance in NUREG-1022, Example C-30, discussed the following situation:

While performing a routine surveillance test of the emergency diesel 1

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 e

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.

f 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)].

l 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) Generic Setpoint Drift Previous guidance in NUREG-1022, Example C-8, discussed the e

following 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 head correction. 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.

Second Draft, 74 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:

l The event is not reportable due to the drift of a single pressure i

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)].

In addition, NUREG-1022, Supplement 1, Question 7.22 provided the following clarification:

l Example C-8 indicates that a setpoint drift problem with a particular switch could be reportable. Would you clarify if setpoint drifts are to be reported if they ar 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.

(13) Maintenance Affecting Only One Train Previous guidance in NUREG-1022, Supplement 1, Question 7.21 posed the following situation:

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 maintenanc.e 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.

Second Draft, 75 NUREG-1022, Rev. 1 i

(14)

Conditions Observed While System Out of Service Previous guidance in NUREG-1022, Supplement 1, Question 7.10 posed the following situation:

f 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.

(15) 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. Although both diesels were operable at the time of surveillance, extended operation without corrective action could have resulted in bearing failure.

j The event is reportable because, although both diesels were operable, there was reasonable doubt that both diesels would have remained operable until they completed their safety function if called upon.

(16)

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 non-containmeat isolation check valve was found lodged in downstream piping. This might have prevented HPCI from functioning if the disc had blocked the line.

HPCI was operable with the disc lodged in the non-blocking position.

The event was caused by fatigue failure of a disc pin.

Second Draft, 76 NUREG-1022, Rev. 1

1 i

The event is reportable because the HPCI could have been prevented from i

performing its safety function.

In addition, the event is reportable if the fatigue failure is indicative of a common-mode failure.

(17) Defective Component Delivered but not Installed I

The previous guidance in NUREG-1022, Supplement 1, Question 7.19, j

discussed the following situati on:

i 1

j How should a plant. report a defective component that was delivered, but l

not installed?

j Answer:

l A single defective component would not generally be reportable (assuming a

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 be reportable.

Engineering judgment is required to determine if the defects could have escaped detection prior to installation and 1

operation. As a minimum, any generic problem may be reported as a voluntary LER.

In addition, such a condition may be reportable under 10 l

CFR Part 21.

J i

(18) Operator Inaction or Wrong Action l

1 i

Dravious guidance in NUREG-1022, Supplement 1, Question 7.25, posed the

)

following situation:

In some systems used to control the release of radioactivity, a detector controls certain equipment.

In other systems, a monitor is present and j

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:

l Yes. The operator may be viewed as a " component" that is an integral, 1

and frequently essential, part of a " system." Thus, if an event or j

condition meets the criterion specified in 50.73 for reporting, it is to be reported regardless of the initiating cause (i.e., whether an 4

j equipment, procedure, or personnel error is involved).

(19) Results of Analysis Previous guidance in NUREG-1022, Supplement 1, Question 7.2, discussed i

the following situation:

l A number of criteria indicate that they apply to actual situations only 1

and not to potential situations identified as a result of analysis; yet, i

Second Draft, 2

77 NUREG-1022, Rev. 1 w

4 l.-

i i

other criteria address."could have." When do-the results of analysis j

have to be reported?.

l Answer:

The results need only to be reported.if > the; applicable criterion -

requires the reporting of conditionsithat-"could have" caused a problem.-

However, others havei a need to know about potential ~ problems that are -

not reportable; thus, such items may.be reported as a voluntary LER.

'i (20) Previous guidance in NUREG-1022, Supplement 1, Question 7 3, discussed:

the following situation:

j Utilities are not required to-analyzeffor' system interactions,,yet;the l

rule requires the reporting of events that,"could have" happened but did -

not. Are we to initiate a design activityjto determinef"could have" system interactions?.

j Answer.

I l

No. Report system interactions that:you find as a. result of ongoing-routine activities (e.g... the analysis-of operating events).

i I

l 3

i 1-i

?!

4 i

i i

a i

1 i

Second Draft, 78 NUREG-1022, Rev. 1 i

l i

3.3.4 Common-Cause Failures of Independent Trains or Channels I

10 CFR 50.72 950.73(a)(2)(vii)

I

[No corresponding Part 50.72 Licensees shall report:

"Any event requirement.]

where a single cause or condition caused at least one independent train or channel to become I

inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:

.i (A)

Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; i

(C) Control the release of 1

radioactive material; or (D) Mitigate the consequences of an accident."

l l

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 caused a component or group of components to become inoperable in redundant or independent portions (i.e.hdfin?the"Esmindskids).iffilldFePhFslmalfshalbniof one or mo trains or channels a safety function. 'Includ h au s ed ibyishch ifist6rsfs s;.h ig h fambisn t Ltempe ratu res Q he a.ts$ f@mjne rgi z at isii l inadequate l preventive 1 maintenance, foil?co'ntaminations f4irlsystemsnincorrect o

.lubricationfuse ofMonquilified;c.ompshent;sionmandfsethfiggQesj@fissW i

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 j

(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 4

system and in Train "B" of another system (i.e., train that is assumed in the safety analysis to be independent) to become inoperable, the event must be i

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.

Second Draft, j

79 NUREG-1022, Rev. 1

Trains L or ?chinsel si fo&eportabil.1 tyl purposefsrenie. fined %sithoieffedsndanti failures. jMany.ns orichan'nelsl designed tOLprovidefprotectionl:against jingle

~

indepen_de'nt? tr'ai engineeredJsafetyjsystemsicontainingfactive;c6mponentsfarei designed;with at:least a1two train;systeme ;Each indepen.dentitrain;in:iaitwos a

tralnisys. tem caninormally satisfy 1allethec:safetyisystemirequireinentsEto"safelf

' hut'.down}thelplantTor satisfy:those feriterialthatc haveitoibejmettfoll6 wing lan s

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 failures and systems interactions. Such failures can be simultaneous failures that occur because of a single I

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

(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 (S0V) failure.

Both units were shut down from 100-percent power, and the SOVs piloting all 16 MSIVs were inspected.

The licensee found that the S0Vs 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 0-rings. As a result, motion of all the S0Vs 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 af 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 [s50.73(a)(2)(vii)(C Second Draft, 80 NUREG-1022, Rev. 1

and D)] and (b) because a single condition could have prevented fulfillment of a safety function [s50.73(a)(2)(v)].

(2)

Marine Growth Causing Emergency Service Water To Become Inoperable (Common-Mode 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 IC 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.

4 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 [650.73(a)(2)(vii)(D)].

i j

(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 -

sual bypass caused these snubbers to malfunction. Tiese snubbers were designed for low probability seismic events.

Numerous previous similar events have been reported by this licensee.

a 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 to fulfill their safety function following a seismic event.

(5)

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-inch stop l

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 i

corroding shut.

Second Draft, 81 NUREG-1022, Rev. 1

The event is reportable b'ecause.a s' ingle' 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 therefore reportable under50.73(a)(2)(vii)(BandD).

i l

e f

Second Draft, i

82 NUREG-1022, Rev. I I

i 3.3.5 Airborne or Liquid Effluent Release s50.72(b)(2)(iv) s50.73(a)(2)(viii) i Licensees shall report:

Licensees shall report:

(A) Any airborne radioactive release (A) Any airborne radioactivity that, when averaged over a time release that, when averaged over a period of 1-hour, results in time period of 1-hour, resulted in concentrations in unrestricted area airborne radionuclide concentrations that exceed 2 times the applicable in an unrestricted area that exceed concentration limits specified in 2 times the applicable concentration Appendix B to gs20.1-20.601, table of the limits specified in Appendix i

II, column 1, of Part 20 of this B, table II of Part 20 of this chapter, or, for licensees chapter, or, for licensees implementing the provisions of implementing the provisions of 9s20.1001-20.2401 of this chapter, ss20.2001-20.2401 of this chapter, 20 times the applicable exceeded 20 times the applicable e

concentration specified in Appendix concentration limits specified in l

B to ss20.1001-20.2401, table 2, Appendix B to s20.101-20-2401, column 1, of Part 20 of this table 2, column 1 of Part 20 to this i

chapter.

j (B) Any liquid effluent release that B) Any liquid effluent release that, when averaged over a time period of when averaged over a time period of 1-hour, exceeds 2 times the limiting 1-hour, exceeded 2 times the l

combined concentration limits in limiting combined concentration l

Appendix B to s620.1-20.601, table limits in Appendix B to @ 20.1-l II, column 2 (see note 1 to Appendix 20.601, table II, column 2 (see note B to @s20.1-20.601), or, for 1 to Appendix B to 9920.1-20.601),

licensees implementing the or, for licensees implementing the provisions of ss20.101-20.2401 of provisions of ss20.1001-20.2401 of j

this chapter, exceeds 20 times the this chapter, exceeds 20 times the i

applicable concentration specified applicable concentration specified in Appendix B to 20.1001-20401, in Appendix B to s 20.1001-20.1401, table 2, column 2, of part 20 of table 2, column 2 of Part 20 of this this chapter, at the point of entry chapter at the point of entry into into the receiving waters (i.e.,

the receiving waters (i.e.,

Continued on next page.

)

i l

I Second Draft, 83 NUREG-1022, Rev. 1

550.72(b)(2)(iv) continued 550.73(a)(2)(viii) continued unrestricted area) for all unrestricted area) for all radionuclides except tritium and radionuclides except tritium and dissolved noble gases.

(Immediate dissolved noble gases.

notifications made under this paragraph also satisfy the 550.73(a)(2)(ix) requirements of paragraphs (a)(2) and (b)(2) of &20.403 of this Reports submitted to the Commission chapter, or, for licensees imple-in accordance with paragraph menting the provisions of 5920.1001-(a)(2)(viii) of this section also 20.2401, s 20.2202 of this chapter.)

meet the effluent release reporting requirements of s20.405(a)(1)(v) of this chapter, or, for licensees implementing the provisions of 9l20.1001-20.2401, s20.2203(a)(3) of this chapter.

If not reported under 650.72(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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of the event.

Licensees are required to submit an LER within 30 days.

Discussion i

Although similar to 10 CFR 20.403 (20.2202) and 20.405 (20.2203), these l

i 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 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , normalize the release to I hour (e.g., if the release lasted 15 minutes, divide by 4).

For releases that lasted more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , use the highest release for any continuous 60-minute period (i.e., comparable to a moving average).

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.

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Second Draft, 84 NUREG-1022, Rev. 1

LERW5fpW(BI)DiitWaitisirWiiiiiditKiOTttie7Qi'sjy@js~sMigigjgg@j@

MelEN!ing iHQtJgiimpyipMtacph; 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.

Examples (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 t

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 times the MPC 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 2 times the applicable concentrations specified in Appendix B to 5920.1-20.601, averaged over 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at the site boundary. An LER is required.

(2)

U1 planned 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 concentration in the unrestricted area, averaged over 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , was estimated by the licensee to be 2.1 times the MPC.

The event was reportable via ENS and LER because the airborne radioactiviu release exceeded 2 times the applicable concentrations specified iri Appendix B to 6920.1-20.601, when averaged over a period of 1-hour.

Second Draft, 85 NUREG-1022, Rev. 1

3.3.6 Contaminated Person Requiring Transport to Offsite Medical Facility s50.72(b)(2)(v) 10 CFR 50.73 Licensees shall report: "Any event

[No corresponding Part 50.73 requiring the transport of a requirement.]

radioactively contaminated person to an offsite medical facility for treatment."

l If not reported under s50.72(a) or (b)(1), licensees are required to notify j

the NRC via the ENS of any such transport as soon as practical and in all cases within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of the event necessitating the offsite transport.

i Discussion The phrase " radioactively contaminated" refers to either radioactively requiredsbutsha(ejgMa_nyjjiitiperso.n.]t.efsurveysfdr# radio {

continiina.ted c1o3hjng_and/or fdfiifsWsNp6) 5tfaljfo'd c9ntamihat_i.ony allonsi j0%;

sino t(bee n icompl et ed { be fore st ran s port (6f}t he s pa rso nlo fp[i t d I

fpg med ica] Et re a tment )3t hs111 c en s e e ys houldima ke3 nl ENSi n o ti fj cati oni See "the i

example.

No LER is required for transporting a radioactively contaminated person to an offsite medical facility for treatment.

1 Example (1)

Radioactively Contaminated Person Transported Offsite for Medical Treatment 4

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, lhe HP technicians returned to the plant with the contaminated protective clothing worn by the worker.

If not reported unaer 950.72(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.

Second Draft, 86 NUREG-1022, Rev. I q

3.3.7 News Release or Other Government Notifications

$50.72(b)(2)(vi) 10 CFR 50.73 Licensees shall report: "Any event

[No corresponding Part 50.73 or situation, related to the health requirement.]

and safety of the public or on-site personnel, or protection of the environment, for which a news release is planned or notification i

to other government agencies has been or will be made.

Such an event may include an on-site fatality or inadvertent release of radioactively contaminated materials."

If not reported under s50.72(a) or (b)(1), licensees are required to notify the NRC via the ENS as soon as practical and in all cases within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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

'cessary 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 e

unusual or abnormal releases of radioactive effluents e

onsite fatality e

Licensees generally do not have to report media and government interactions ublic or unless they are related to the radiological health and safety of the p!NRCid6s~s n6t3Ehpfi1]Q5 pdm 63((ilh(6@Qp(EF3hMI(f[tEfijnMI

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onsite personnel, or protection of the environment.

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1 i

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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, tha "'C needs accurate, detailed information in a timely manner regardi i situations. The NRC should be aware of information that is I

availabh.

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 ordt to exclude administrative matters or those events of no safety significance If a particular effluent release has safety significance. or is expected to generate public, media, or other agency attention as a result of being unusual or abnormal, then an immediate notification to the NRC would be warranted.

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, licen should make an ENS notification before issuing a press release becaure new. nedia representatives will usually contact the NRC public affairs officer shortly after its issuance for verification, explanation, or interpretation of the facts.

l 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 950.72(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.

l Second Draft, 88 NUREG-1022, Rev. 1 l

Routine reports to a local, State, or Federal agency that do not involve an event or situation, related to the health and safety of the public or on-site j

personnel, or protection of the environment needs to be reported to the NRC only when that matter get escalated to a " news release" of a " situation".

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.

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 l

Protection Program, a licensee representative was interviewed on local television and radio stations. The licensee notified State officials and the NRC resident inspector.

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(3)

Local Government Notification The licensee contacted the local fire department when a small trash bag in the containment building was ignited by welding sparks. The fire was extinguished within 4 minutes of its discovery and did not result in any damage to plant equipment. The local fire department responded but did not enter the plant site because the fire had been extinguished.

An ENS notification is needed because the local fire department was notified in response to an event related to the health and safety of on-site personnel.

(4)

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 4

Second Draft, 89 NUREG-1022, Rev. 1 4

the licensee of the problem because of the concern of the public for their radiological health and safety.

(5)

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 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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.

(6)

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.

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.

(7)

Routine Reports Regarding Endangered Species The licensee notified the U.S. Fish & Wildlife Service and a State agency that an endangered species of sea curtle was found in their circulating water structure trash bar.

No press release was issued.

An ENS notification is not needed because routine environmental reports of this nature to State and Federal agencies do not involve an event or situation, related to the health and safety of the public or on-site personnel, or protection of the environment.

(8)

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 repa' ired.

Second Draft,.

90 NUREG-1022, Rev. 1 l

l

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i A licensee notified the Federal Aviation Agency that it removed part of j

its auxiliary boiler stack aviation lighting from service to replace a j

faulty relay.

I 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 l

up by removing the gravel and dirt.

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91 NUREG-1022, Rev. 1 4

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3.3.8 Spent Fuel Storage Cask Notifications 9 50.72(b) (2) (vii) 10 CFR 50.73 Licensees shall report:

"Any (No corresponding Part 50.73 instance of:

requirement.]

(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 d general license issued under 972.210 of this chapter.

A followup written report is required by s72.216(b) of this chapter including a description of l the means employed to repair any defects or damage and prevent recurrence, using instructions in s72.4, within 30 days of the report submitted in paragraph (a). A copy of the written report must be sent to the administrator of the l

appropriate Nuclear Regulatory Commission regional office shown in Appendix D to part 20 of this chapter."

If not reported under s50.72(a) or (b)(1), licensees are required to repcrt any such instances to the NRC via the ENS as soon as practical, and in ali l

cases witnin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . A followup written report is required by 972.216(b) l within 30 days.

l 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 s50.72, then as indicated in s21.2(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.)

Second Draft, 92 NUREG-1022, Rev. I 1

l

3.4 Followup Notification This section addresses 550.i2(c), " Followup Notification." These notifications are in addition to making the required initial telephone notifications under 650.72(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, l

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1 Second Dreft, 93 NUREG-10224, Rev. 1

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!l 3.4.1 Followup Reports I

$50.72(c) 10 CFR 50.73 With respect to the telephone

[No corresponding Part 50.73 notifications mada under paragraphs (a) requirement.]

and (b) of this section, in addition to a

)

making the required initial notification, i

each licensee shall, during the course of the event:

(1) immediately report l

(i) any further degradation in the level of safety of the i

plant or other worsening i

plant conditions, including i

those that require the declarution of any of the i

Emergency Classes, if such a l

declaration has not been previously made, or 1

(ii) any change from one Emergency Class to another, or i

(iii) a termination of j

the Emergency Class i

(2) immediately report

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(1) the results of ensuing evaluations or assessments of j

plant conditions, j

(ii) the effectiveness of response or protective measures taken, and (iii) information related to plant i

behavior that is not understood (3)

Maintain an open, continuous.

l communication channel with the NRC l

Operations Center upon request by the NRC.

Di scus sio_!1 These criteria are intended to provide the NRC wi'.ri timely notification when i

an event becomes more serious or additional information or new analyses

)

clarify an event. They also permit the NRC to vintain a continuous l

communications channel because of the need for sontinuing followup information or because of telecommunications problems.

With regard to the open, continuous communications channel, licensees have a i

responsibility to provide enough onshift personnel, knowledgeable about plant j

operations and emergency plan implementation, to enable timely, accurate, and t

i Second Draft,

}

94 NUREG-1022, Rev. 1 l

i

i reliable reporting of operating events without interfering with plant -

operation as discussed in the Statement of. Considerations for the rule and i

Information Notice 85-80, " Timely Declaration of an Emergency Class, Implementation of an Emergency Plan, and Emergency Notifications."

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l 4 EMERGENCY NOTIFICATION SYSTEM REPORTING l

This section describes the ENS referenced in 10 CFR S0.72 and provides general and spec fic guidelines for ENS reporting.

4.1 Emgroency Notification System The NRC Operations Center is the nucleus of the ENS and has the capability to I

handle any emergency communication need. The NRC's response to both.

emergencies and non-emergencies is coordinated in this communication center.

The key NRC emergency communications personnel, the emergency officer (E0),

regional duty officer (RD0), and the headquarters operations officer (H00),

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 (E0F). A separate ENS line is installed at E0F's which are not onsite.

The ENS is part of the Federal Telecommunications System (FTS).

This i

FTS ENS replaces the dedicated ENS ringdown telephones used previously and provides 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 E0F and, like the E E they are now part of the FTS.

(3)

Testing l

As indicated in Information Hotice 86-97, " Emergency Communications System," dated November 28, 1986, licensees should initiate monthly tests of the ENS telephores 20 the NRC Operations Center.

(It is not necessary to test the connection to the NRC regional office because this connection is made by the NRC headquarters operations center.)

In addition, licensees should maintain a record of monthly tests for their entire emergency communications facilities.

Second Draft, 96 NUREG-1022, Rev. 1

ThsMissht?Eh'ahdisisFZt;6TFTS'!20_00JF(U]pEsnpijdTEspsfiijsdilh}tistidd indicate?that)the?following;testyprocedures1:are;morefappropriatefand theg mah be)used M 1jsulofsthejpro~cedures[ discussed $1_nyJNJ 6}97j Licen~seesVay~ac'c'omplish' the onthly"te~st"of ENS bf"pla"cin'"a singis m

9 ball from one 'of the ENS extensions ~ to the NRC' Operations Center. When the: connection is established by the licensee all ENS extensions should be taken off-hook and a voice check performed from each ENS' extension.1 This > verifies that local switching from the, site to the<FTS 2000 system is functioning. properly and that acceptable'servtce is available.(no~ ~

signal degradation) on the ENS line.' A return call from the NRC Operations Center should be requested before_ breaking the connectioij ^an.d returning' all ENS extensions on-hook. When' n e~ connection is re-established by the NRC Operations Center all ENS extensions.should;agaih be'taken off-hook and a voice check performed from each ENS extension.,'

This verifies that local switching from the NRC to the FTS 2000 system is functioning properly and that acceptable ' service is 'available~(no'"

signal degradation) on the' ENS line,J finally, to demonstrate'that each.

of the ENS extensions may initiate a call. the licensee may place a call ~

in-house on an ENS extension to another FTS line,such as one of the Counterpart Links. This conipletes the verification that' the ENS lide functior.s properly and that the FTS 2000 system is accessiblec See

Se~ction 3.2.7 for guidance onreportin'g' of any ENS"pf6blsnis.

Licensees should also document and maintain a record of the monthly tests of the HPN, Reactor Safety Counterpart Link (RSCL), Protection-Measures Counterpart Link (PMCL), Management Counterpart Link (MCL) and the Local Area Network (LAN) Access. As indicated in Information Notice 86-97, all HPN instruments may be tested by placing local calls (to or from a Counterpart Link). No call to the NRC Operations Center is necessery._Tssfi;hj@fsthiiCidhtsfiaFMihks{shilbiliH6hiplJs;hidiby,,e Mi6ljis$E s ti ng;th e(Loc al @ Area ! Ne two rkk(LAN)?Acce(C66nt e rp hi nglalcon n ec ti onlt o gnd/ from j a 3 d i f fe F6n tsspinvolvesttemporaril, fem #spho ne! a nd les tabl i shi ng{[ con nect i_on s d housek

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Establishing lasconnectionteither tofornfrom8thelNRC10perationgCFdtEh;j Testing of the Emergency Response Data System (ERDS) Channel is performed quarterly unless otherwise set by NRC based on demonstrated system performance. Details are provided in Generic Letter 93-01, Emergency Data System Test Program, dated March 3, 1993.

Under the current arrangements the NRC furnishes FTS 2000 service up to the demarcation distribution frame or "demark" and the telephone sets.

The NRC also furnishes the modem for ERDS. As discussed in Section 3.2.7 6blimsMi tENRCEf0FhishsdiE4Ulpiiis6tish6uld? bbifhpiftsditbiths NRCf0p{e#sti6n s7 Ce nt e s soith atfFeia i'rs tc an t

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Second Draft, 97 NUREG-1022, Rev. 1

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licensee support of the FTS 2000 is discussed in Generic Letter 91-14:

Emergency Telecommunications, dated December 23,1991.)

1 (4)

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.

(5)

Facsimile Transmission (Fax)

Licensees occasionally fax an event notification into the NRC Operations Center on a commercial telephone line before making an ENS notification.

2 However, 950.72 requires that licensees notify the NRC Operations Center via the ENS; therefore, licensees also must make an ENS. notification.

4.2 General ENS Reportina 4.2.1 Reporting Timeliness The required timing for ENS reporting is spelled out in 9950.72(a)(3),(b)(1),

(b)(2), (c)(1), (c)(2), and in the Statements of Considerations, as "immediate" 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 the NRC may be interested in. The NRC responds to any voluntary

~

1 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 l

the ENS notification to a required notification under the appropriate 10 CFR 50.72 reporting criterion.

4.2.3 ENS Notification Retraction 4

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.

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 Notific:: tion Second Draft, 98 NUREG-1022, Rev. 1 i

Worksheets. The worksheet provides the usual order of questions and discussion for easier communication and its use often enables a licensee to i

j prepare answers for a more clear and complete notification. A clear ENS notification helps the H00 to understand the safety significance of the event.

Licensees may obtain an event number and notification time from the H00 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 j

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 l

avoid using local jargon for plant components, areas, operations, and the like so that the H00 can quickly understand the situation and have fewer questions.-

In addition, others not familiar with the plant can more readily understand the situation.

4.3 Tvoical ENS Reportina 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 H00 needs to understand the safety significance of each event to brief NRC mn.agement or initiate an NRC response.

The H00 will be primarily concerned about the safety significance of the event, the current condition of the plant, and the possible near-term effects the evert could have on plant safety. The H00 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 l

description of the event, the H00 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 H00 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.

1 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 2

protect the health and safety of the public.

When required, the NRC supports the licensee with technical analysis and coordinates logistics support.

The l

Second Draft, 99 NUREG-1022, Rev. 1

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 l

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 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 I hour or. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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-l the most serious events occurring at operating nuclear power plants without i

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).

l l

Second Draft, 100 NUREG-1022, Rev.-1

s 5

LICENSEE EVENT REPORTS This section discusses the guidelines for preparing and submitting LERs.

Section 5.1 addresses administrativa requirements and provides guidelines for submittals; 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 Reportina Guidelines This section addresses administrative requirements and provides guidelines for suismitta; 3.

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 550.73(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 550.4."

3 i

An LER is to be submitted (mailed) within 30 days of the discovery date.

If.a 30-day period ends on a 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 4

the report, the situation should be discussed with the appropriate NRC regional office.

See Section 2.11 for further discussion of discovery date.

1 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 (INP0) in their distribution.

Multiple LERs can be forwarded -

by one forwarding letter.

Second Draft, 101 NUREG-1022, Rev. 1

Canse11ations of LERs submitted Eshabid'be isade'bfflettsR"Thelbasss?foMthe cancellation.should be explained so thatithe>staffican understandiardireview o

thhseasons supporting the'.determinationt ;The noticeto'f) cancellation!Wil19bs fired and stored with the LER and.acknoWledgementsjmadesi@!;rbus[ automat.ed datafsystems.

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 650.73(f)

"Upon written request from a licensee including adequate justification or at the initiation of the NRC staff, tie 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 Th'e' Commission encourages : voluntary. LERs1rather;thansinforniattorN1ettsrs70W10 CFRL50.9' verbal l reports to report op' r'ational sysnts that"do not'ineet the e

criteria contained'in 10 CFR 50.73. The LER format is preferable because of the established procedures for distribution and entry into computerized data files.

The NRC recognizes that the number of LERs is not in itself an accurate or aopropriate measure to judge a plant's safety performance.

Also see Section 2.9.

Because not all requirements of s50.73(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.

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 Second Draft, 102 NUREG-1022, Rev. 1

immediately below the block. Also give a sequential LER number to the voluntary report as noted in Section 5.2.4(5).

5,1.6 Svoplemental Information and Revised LERs

- =.

950.73(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 650.4, 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 J

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 3revious report in the com In addition indicate in the text l@jpegMBgm@gg}bb5NINN!$puter file.Mk )NNMIMNfN dNIM h

s 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 cours% significance, implications,-or consequences of the event or if it results in substantial changes in the corrective action planned by the licensee.

l 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 i

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.

I? a criterion for reportability was checked in Item 11' of NRC Form 266 and later it was determined that other requirements also pertain, a revised LER should be submitted.

When a voluntary LER is submitted and later it was Second Draft, 103 NUREG-1022, Rev. I f

b

determined that the event was required to be reported, submit a revised LER to identify this fact.

5.1.7 Special Reports i

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 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 550.73(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 use6 to submit a special report, use a report number from the sequence used for LERs.

1 If an event is reportable both undar 10 CFR 50.73 and as a special report, 1

cneck 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 4

on the reportable situation, j

5.1.8 Appendix J Reports (Containment Leak Rate Test Reports)

A licensee must perform containment integrated and local leak rate testing and i

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 4

report by reference, if desired. The LER should address only the reportable situation, not the entire leak rate test.

1 5.1.9 10 CFR Part 21 Reports WiCFRQaFtT21p"l i cen see stofg( DifeEfs [andJ6hj6i651]ahdeNsFsmiddiCdsid Rspoft}hp0f

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~~

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 wordin

" Abstract") and Item 17

'("Tsxt")7thiuldistifEthitEtheWep6Ft?g in Item 16 (PaFts cohstitutis W WE$de fe c tMappl Li c abl e% {othdrffsci11 t.is sVat WTmul t i fbni t91 inOs LER;may;belused@yilndicatingithero.therdnvolvedhfacilitiesiinjItem MRQormy Second Draft, 104 NUREG-1022, Rev. 1

1 5.1.10 10 CFR 73.71 Reports

)

Submit events or conditions that are reportable under 10 CFR 73.71 using the j

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 safeauards 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 requirementsof@73.21(glmustbemetwhentransmittingsafeguards information. F6FidditiossR(Jiildslinds Riid11t'aFFGI61 del 5;62 MReV.i siob?1MRsp;id6M0JCFR173MIWsE6rtTBW 1987RNUREGil304@Repbrti

[gMyQ1103lf[R epoIt j ngM.n'g ?6 f(S a feg n ard sf Ev6nt sid FebWiafR1988j]at d afeguard g Ejentsl & % f M & f M 9 R If the LER contains proprietary information, mark Item 17 of the LER form.

Include proprietary information only in the narrative and not in the abstract.

In addition, indicate clearly in the narrative the information that is 1[ppri,e.tary._ Jins 11y@hsIFe@fsifidhtsT6qs@((p)[niGitybE@@hsii PIan q @ lggypppgrieta n dpfgrmatio g 5.1.11 Availability of LER Forms The NRC will provide LER forms (i.e., NRC Forms 366, 366A, and 3668) free of charge. Copies may be obtained by writing to the NRC Publication Services i

Section, U.S. Nuclear Regulatory Commission, Washington, DC 20555. A facsimile of an LER form may be-used to facilitate word processing, but the size and general format of the LER form should not be significantly altered.

5.2 LER Content Reauirements and Preparation Guidance 3

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 whpn appropriate. This?ijstJ.bnysi s6?FFby]'dssjthigtspijt5

.6poWedjf5Iprepapingkan;LERSjhfiUdingtaisuggestedfordemforfpreparing[bs,E f

@strittWdsdidelon?att i tl eit ha Rcipt&j o f! theleventMtext]

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sntheisi hientH5iiEfillMnnhsj otWfisi dsjf3hsMohi)$ghi fi shntfelussn ts f 6fsth'sMske g6thQER{ forms In 1986, the NRC decided to ute an optical character reader (OCR) to read LER abstracts into NRC LER data bases (!F 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.)

Second Draft, 105 NUREG-1022, Rev. 1

Type Styles:

Prestige Elite (12 pitch)

Letter Gothic.(12 pitch)

OCR-B (12 pitch)

Courier 12 (12 pitch)

Elite (12 pitch)

Courier 10 (10 pitch)

OCR-A (10 pitch)

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 letter-quality) printer (e.g., daisy wheel, laser, ink-jet).

It is suggested that output have an uneven right margin (i.e., we suggest that you not right justify output).

It is suggested that text of the abstract be kept at least 1/2-inch inside the border on all sides of the area designated for the abstract on the LER' form.

It is suggested that you do not underscore, use bold print, use. Italic print style, end any lines with a hyphen or 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 10'6 could be expressed as 4.2E-6, 0.0000042, or 4.2 x 10(-6).

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Whsri;lfepo Fti hphE tualj70 Fp6 t e h ti hlf;;fa l l bfesT6 ffE6inp'o neht'5i st;Fui risifo 6sfit emu describsjnyfred.Undantidmponentsstrains.Jdr/systemsithatsere :avai,lable: land operable

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(1)

Format 550.73(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.

Although s50.73(b) defines the information that should be included, it is not intended as an outline of the text format. After the narrative is written, however, review the appropriate sections of s50.73(b) to make sure that applicable subjects have been adequately addressed.

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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).

Second Draft, 107 NUREG-1022, Rev. 1

i 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 engineered safety feature (ESF) 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 random 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. There is'no need to provide redundant information or unimportant details, but it is necessary to discuss the performance and status of ESF 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 and large number of reviewers that read each LER.

In cases where the information in the SAR may not be.sufficiently detailed or up to date, add the necessary information in the LER.

It is not necessary to-include excessive technical detail in the-narrative that would detract from readability of the report.

Ensure, however, that each applicable component's safety-significant effect on the event or condition is clearly and completely described.

(2)

Specific Information 850.73(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.

l 850.73(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.

Second Draft, 108 NUREG-1022, Rev. 1

i 550.73(b)(2)(ii)(c)

The narrative description must include:

" Dates and approximate times of occurrences."

Provide sufficient times and dates in the narrative to capture the time sequence of the event. The event date is generally the day on which the event occurred; if the event date is not known or is uncertain, the event date can be the discovery date. 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.

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ectsh 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; and stable conditions achieved).

Include an estimate of the time and date of failure of systems, components, or trains if different from the time and date of discovery.

For example, if an ESF actuated on January 15, 1991, but the actuation was not discovered until a review of the sequence-of-events printout on January 30, i

the event date should be January 15. However, if a licensee discovered on January 15, 1991, that a design error occurred some time in 1982, then the event date should be January 15, 1991.

Components such as valves and snubbers may be tested over a period of several weeks. During this period, a number of inoperative similar components may be discovered.

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 1

first failure are 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 i

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 informatior. 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).

Second Draft, 109 NUREG-1022, Rev. 1 4

4

550.73(b)(2)(ii)(D)

The narrative description must include:

"The cause of each component or system failure or personnel error, if known."

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550.73(b)(2)(ii)(E)

The narrative description must include:

"The failure mode, mechanism, and effect of each failed component, if=known."

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i 1

$50.73(b)(2)(ii)(F)

)

The narrative description must include:

"The Energy Industry.

l 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.

j (2)

IEEE Std 803-1983 has been approved for 7

incorporation by reference by the Director of the Federal Register.

1 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 i

Institute of Electrical and Electronics Engineers, 345 East 47th Street, i

New York, NY 10017.

IEEE Std 803-1983 is available for inspection at the NRL's Technical Library, which is located in the Phillips Building, 7920 Norfolk Avenue, Bethesda, Maryland; and at the Office of the Federal Register, 1100 L Street, NW, Washington, DC."

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),

4 the system code (e.g., B0) 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.

i j

Whenever an uncertainty arises concerning the interpretation of a system boundary, for those systems included in the nuclear plant reliability data i

system (NPRDS) reportable scope, the boundary should be defined consistent with the comparable system descriptions and interpretations contained in the j

NPRDS Reportable System and Component Scope Manual.

550.73(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 i

were also affected."

i t

i~

111 NUREG-1022, Rev. 1 Second Draft, i

i

No further explanation is necessary.

550.73(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."

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The narrative description must include:

"The method of discovery of each component or system failure or procedural error."

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I Second Draft, 112 NUREG-1022, Rev. 1

650.73(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 l

\\

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-i licensed operator, other utility personnel)."

j Human performance often influences the outcome of nuclear power plant events.

Human error is known to contribute to more than half of the LERs as discussed in previous guidance. 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 operation occurrences.

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)(d) require a description of (1) operator actions that affected the course of the event and (2) for each personnel error, additional specific information detailed in the rule; for example, whether the error was a cognitive error or a procedural error.

Second Draft, 113 NUREG-1022, Rev. I

$50.73(b)(2)(ii)(K)

The narrative description must include:

" Automatically and manually initiated safety system responses."

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$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."

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(3)

Assessment of Safety Conseauences

$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.

l Second Draft, l

114 NUREG-1022, Rev. I

Reasonable and credible alternative conditions may include normal plant I

operating conditions, potential accident conditions, or additional component' failures, depending on the event. Normal alternative operating ' conditions and i

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 950.73(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."

1 Discuss all ' corrective 'setion's' 'or' enhancement ~s'thit'tes01ted fr' m'thi ivent.,

o The narrative should include the corrective actions that were tracked'by the licensee's internal corrective action, system.

Include when the corrective action was or will be implemented. The term " corrective actionst includes both the actions to restore the system or component to service and the actions

,to prevent recurrence. Discuss re) air or' replacement actions as well as '"' '

actions that will reduce the proba)ility of a similar event occurring in"ths future. For example, "the pump was repaired and,a discussion of the event wa5 included in the training lectures." Another example, ;"al though no ~~~~"~~ ~'

' modification to the instrument was deemed necessary,"a caution nots ~wss"placed in the calibration procedure for the instrument before the step ^'in which the

"'^

^ ~ ~ ~ " ~ ~

~ ' ' " '

event was, initiated "'

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.

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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.)

Second Draft, 115 NUREG-1022, Rev. 1

(5)

Previous Occurrences 950.73(b)(5) i 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' 1 ice'dse'esh'6uld usi'En'ginbsFingjddg~m'd6t"ts~dscids' how~'fsY' back"in" time, t'o go to present'a reasonably complete picture of the current problem.' The_

intent is,to be able to see'a pattern in recurring events, rather than to get a complete 10: or 20-year history of the system. 'If the event was a high-frequency, type of event,'2 years back may. be more than' sufficient.

If' corrective actions keep changing'and the same type of'Went keeps ^ occurring, then the, root cause has not,been addressed /

' " ' ~ ~

~'

nElsds?tFsXERinsiiibsF(j]?i[Milly?33f5^fiesiEiliffEll^iFsisHYsh HU635FsViEEs

'similassVentsi6centredas# state 1MThilyle&sdayihtfsass6thsfeff1s'ieh sand pffestjieKbjnpfd@isilfic$effeQppEo66jsi~If anfiW1liF~ssif ER retrospect, were s ant in relallon t'6 the subject event, discuss why prior corrective action did not prevent recurrence.

(6)

LER Text Continuation Sheet (NRC Form 366A)

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 1/2 x 11).

Second Draft, 116 NUREG-1022, Rev. 1

5.2.2 Abst act (NRC Form 366, Item 16) f50.73(b)(1)

The LER shall contain:

"A brief abstract describing the major occurrences during the event, including all component or system failures that J

contributed to the event and significant corrective action taken or planned to prevent recurrence."

l 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.

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.

4 It is acceptable to describe the entire event in the abstract space. However, the description of the event should be sufficiently detailed so that a knowledgeable reader can understand the complete event.

Few reportable events eill be so simplistic that they can be adequately described in 1400 4

characters.

The abstract is generally included in the LER data base to give users a brief 3

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.

i As noted in Section 5.1.10, do not include safequards, security, or proprietary information in the abstract.

5.2.3 Event Title (NRC Form 366, Item 4) 1 The title should include a concise description of the principal )Voblem of issue associated'with' the event, the7oot cause',' the re' Ult'(whi,the ' event 'sas

^

i

' equired to be reported), and the link between them,,1f 'possible ' It is often r

easier to form the title after writing'the assessment >of the event because the information,is clearly at hand.

' ^ ' ~ ' ~ ~ ' ' ~ ~ ' ^

'~

robt'cidss'a,The" titre !Re~actdF m

" Licensee Event Report" should not be used as a title.

nd'the link betwee6 Trip,is' con' side' red 16adequat'e, bec'asie thir the root cause and,the result are missing. The< title " Personnel Error,Causes Reactor Trip" is considered inadequate because'of the innumerable ways in' ~'

Which a person could cause a reactor trip, 'e," Technician Inadvertently Injicthd Signal _ Resulting in a Reactor Trip" would,b a better titleg " '

' ' ~ '

4 Second Draft, 117 NUREG-1022, Rev. 1

R i

5.2.4 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)

Paae 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, i

number the pages consecutively beginning with page 2.

The front side of the two-sided LER form, including the abstract and other data is pre-numbered on the form as page 1 of __ ; the back side of the form actually starts page 2 and needs to be numbered.

(4)

Event Date (NRC Form 366. Item 5)

Enter the date on which the event occurred in the six spaces provided. There are two spaces for the month, two for the day, and two for the year, in that i

order. Use leading zeros in the first and third spaces when appropriate.

For example, June 1, 1987, would be properly entered as 060187. Use the discovery date if the event date can not be clearly defined.

I (5)

Reoort Number (NRC Form 366. Item 6)

The LER number consists of three parts: (a) the last two 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 3

submittal was the 1st revision to the original LER for that event.

Event Sequential Revision j

Year Report Number Number l

91 045 01 Event Year:

Enter the last two digits of the year in which the event occurred. For example, for events occurring in 1991 enter 91 in the spaces provided.

Seouential Report Number: As each reportable event is reported for a unit during the year, it is assigned a sequential number.

For example, for the Second Draft, 118 NUREG-1022, Rev. 1

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.

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.

The sequential number should begin with 001 for the first event that e

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 05000XXX 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).

(6)

Report Date (NRC Form 366. Itgm 7)

Enter the date the ilR is submitted to the NRC in the six spaces provided, as described in Section 5.2.4(4) above.

(7)

Other Faciiities (NRC Form 366. Item 8)

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 rcfer 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 usn of the lowest number should only apply.

l 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 Second Draft, 119 NUREG-1022, Rev. 1-

directly affected by the event (e.g., the event included shared components, the LER described a tornado that threatened both units of a two-unit plant).

(8)

Operatina Mode (NRC Form 366. Item 9)

Enter the operating mrde of the unit at the time of the event as defined in e

the plant's technica' specifications in the single space provided.

For plants i

that have operating modes such as hot 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.

(9)

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.

(10) Reportina Renuirements (NRC Form 366. Item 11)

Check one or more blocks according to i.he repr,rting requirements that apply to the event.

A single event can meet more than one re' sorting criterion.

For l

example: if as a result of sabotage, reportable under f73.71(b), a safety l

system failed to function, reportable under $50.73(a)(2)(v), and the net result was a release of radioactive matarsal in a restricted area that I

exceeded the applicable license limit, reportable under s20.405(a)(1)(iii),

prepare a single LER ar.d check the three boxes for paragraphs 73.71(b),

50.73(a)(2)(v), and 20.405(a)(1)(111).

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 caso of attempted sabotage (973.71(b)) that does not 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.

(11)

Licensee Contact (NRC Form 366. Item 12) 550.73(b)(6)

The LER snall contain:

"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."

Second Draft, 120 NURrG-1022, Rev. 1

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.

(12) Component Failures (NRC Form 366. Item 13).

Enter the appropriate data for each component failkre 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.

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 iten 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 cloraly describes the root cause of the failure.

Cause Code Classification and Definition A

Personnel Error 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 Desian. Manufacturina. 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.

Second Draft, 121 NUREG-1022, Rev. I

E Manaaement/0uality Assurance Deficiency is assigned to failures caused by inadequate management oversight'or management systems 1

(e.g., major breakdowns in the licensee's administrative controls, prevent'ive maintenance program, surveillance program, or quality assurance controls, inadequate root cause determination, inadequate corrective action).

X Other is assigned to failures for which the proximate cause cannot be identified or which cannot be assigned to one of the other classifications.

i 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.

i component:

Enter the applicable component coda from IEEE Standard 803A-1983, "IEEE Recommended Practice for Unique Identification in i

Power Plants and Related Facilities - Component Function Identifiers."

Component Manufacturer: Enter the four cht.racter alphanumeric reference code.

Chapter 18 of the'NPROS Reporting Guidance Manual 4

describes how to access a computerizr:d listing of manufacturer codes maintained on INP0's computer. Designate manufacturers that are not included in the list as X999

)

.Rgp;rtable to NPRCS:

Enter a "Y" if the failure is reportable to NPRDS and an "N" if it is not reportable.

Inc11de in the LER text and in item 13 of the LER Form any compt nent failure involved in the event, not just components with'n the scope of NPRDS or EIIS.

[3a Qure Continuation Sheet (NRC form 366til:

If more than four failures need to be coded, use one or more of the failure continuation sheets (NRC Form 3668). Code the entries in Items 1, 2, 3, and 6 of tre failure continustfon ;heet to match ent. ries 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 r.ny failure continuation sheets after any text continuation sheets and include those sheets in the total nunbar of pages for the LER.

(13)

Supplemental Report (NRC Form 366. Item 14)

Check the "Yes" block if ';he licensee plans to submit a followup report.

For example, if a failed component had been returned to the manufacturer for Second Draft, 122 NUREG-1022, Rev. 1

MRC FOHbJ66 U.S. NUCLE.AR REGULATOR 1 COMMISSION APPROVED BY OMB NO. 3150 0104 Dem EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WrrN THis UCENSEE VEM=kt7 g g ny gg ng INF0FWAT10N COLL.ECTION REQUEST: 90.0 HRS.

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RI g6 nf COMMENTS REGARDING BURDEN EST1 MATE TO THE INFORMATION AND RECOFOS MANAGEMENT BRANCH lWNES 7744), U.S. NUCLEAR REGULATONY COMMISSION WASHINGTON, DC 20LS0001. AND TO THE PAPEPMORK REDUCTION PROJECT A190410s), OFFICE OF (See reverse for required nuriber Of digits / characters f0f each block)

MANAGEMENT AND BUDGET, WASHINGTON, DC 20603 PAC 3UTY NAME p)

DOCKET NunsetR m PAGE W 05000 1 OF nTts es)

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EVENT DATE (5)

LER NUMBER (6 REPORT DATE (7)

OTHER FACILITIES INVOLVED (G) sEoVEm e SION F AOUTV MME DOCMT NUMWER lMONTN oAv vfAR vfAR D^Y NuMeER NuusEsi 05000

~

FAcuTY N/Mti DOCGT NUMBER l

05000 OPERATING THIS REPORT IS SUBMrTTED PURSUANT TO THE REQUIREMENTS bF 10 CFR k ICheck *>ne or more)(11) 4 MODE (t) 20.402(b) 20 405(c) 50.73(a)(2)(iv) 73 71(b) j POWER 20.405(a)(1)(1) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)

LEVEL (10) 20.405(a)(1)(a) 50.36(c)(2) 50.73(a)(2)(vis) 1 Gir,cH i

20.405(a)(1)(in) 50.73(a)(2)(i) 50.73(a)(2)(vio)(A) e "= m; a Tui, NRC lai T* * *$

20.405(a)(1)(ev) 50.73(a)(2)(n) 50.73(aH2)(viu)(B)

,, m y s 2 ^' -

20.405(a)(1)(v)

%.73(a)(2)(in) 50.73(a)(2)(x)

LICENSEE CO_F, TACT FOR THIS LER (12)

J COMPLETE ONE LINE FOR EACH Cf MPONENT FAILURE DESCRIBED IN THIS REPORT (13) s CAU8E Sv37EM COMPONENT MANUFACTURER I

NPRD y

CAUSE SYSTEM COMPONENT MANUFACTURER 1

Y 4

SUPPLEMENTAL REFORT EXPECTED (14) i EXPECTED MONN DAV YEAR SUBMISSION p y, esmpone EXPECTED sUBW8SION DATE)

DATE (15)

ABSTRACT (Umst to 1400 spaces, i.e., apprOximately 15 single spaced typewntten lines) (16) i i

1

- - - 'MI Second Draft, i

124 NUREG-1022, Rev. 1

additional testing and the results of the test were not yet available wne: the

~

LER was submitted, a followup report would be submitted.

(14) Expected Submission Date of Supolemental Report (NRC Form 366. Item 1 Enter the expected date of submission of the supplemental LER, if-applicable. ~ ] $Y T W A $pected submission date r date format. The ex See Section 5.2. for the $?b ~ 1 i i 1 l Second Draft, 123 NUREG-1022, Rev. I

NRC FORM 384A U.S. NUCLEAR REGULATORY CC-^ - -B --. rWD BY 0805 NO. 31eM104 Dem EXPT #ES 5/31/96 est=Atuo sunneN pea esposes to Cowtv wm ws LICENSEE EVENT REPORT (LER) "o " '"." O " "no". y' E 'o O w k"" E S c. u AN asm Mmaesuem ammca mees nie.v s sa;CosAa TEXT CONTifiUATION MEOULATORY NW. WAspeNOTON, DC90004 000t. AND TO THE PAPGAWORK REDUCTION PRQJECT pito.0104. OFFCE 08 MANAGEMEW AND BuceET.waaPSESTON. DC 30003. 7ActuTV NAme su accetET seuteten sap Lan tsuengem est PAe4 sai SEQutNTIAL fEA8 SON HuWeEh HUMBER 05000 0F a, sa - a mC on m on 1 1 1 l Second Draft, 125 NUREG-1022, Rev. I

NRC FORM 3445 U.S. NUCLEAR LEGULATOwY COMMISSION APPROVED BV OMB NO. 31504104 Sea. EXPIRES 5/31/95 EsTWATED SUROEN PER RESPONSE TO COMPLY Wm4 TH3 LICENSER,. EVENT REPORT (LER) NF0FEAAT10N COLLECTION FEQUEsT: 50.0 HRS FOmfARD COMuENT:REaA,oNo suRoE Envium 70 w woRuAnON FAILURE CONTINUATION AN REC R S MW EMENT SRANCH WN88 Wl, U S. MW RiiGULATON COhMSSON, WA&eM3 TON, DC 30581H1001. AND TO THE PAPERWORK REDUCTION PFQJECT Q1504104), OFFICE OF MANAGEMENT AND SUDGET, WA&1NOTON, DC 20503. PACluTY MAMs tip OOCKET NuMSER (2) LER NUMdHER le) PAGE (3) SEQUENTIAL REYSON yp NUMSER NUMBER 05000 OF COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) l CAUSE SYSTEM COMPONEC MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTUFER "m TO PRDS l l 1 l l 1 1 I l 1 l l NRC,0,w m.. Second Draft, 126 NUREG-1022, Rev. 1

2 .i i l 4-1 4 1 3 1 i APPENDIX A 1i l HISTORICAL PERSPECTIVE ON EVENT REPORTING i s-l 4 i 1 i 1 j 6 1 4 r j a 1 1 i 5 1 4 + t 4 i l. f i i t - 3 f' i i

o 4

e i i i 5 i l j Second Draft NUREG-1022,-Rev. I i 1 l i 4 T .I

Oriain 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 (10ER) system. The 10ER system would 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 prc;;o:ed rulemaking concerning the 10ER system was publishad on January 15, 1981 (46 FR 3541). On June 8,1981, the Institute of Nuclear Power Operations (INP0) 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 j data and further development of the 10ER system was discontinued. 1 On May 6,1582, the NRC published a notice of proposed rulenaking in the federal Regfrter (47 FR 19543) that would modify and codify the existing LER system. On July 26, 1983, after consideration of public' conments, 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 1s a vital adjunct to 10 CFR 50.73 for component data. 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 do'cumentation 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 Reg / ster (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-deptn tcilowup report under 10 CFR 50.73. NRC Workshops and Event _Reportina Guidelines In September 1983, the NRC staff published NUREG-1022, " Licensee Event Reporting System," to urovide supporting information and guidelines to persons responsible for the 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 Second Draft, A-1 NUREG-1022, Rev. 1

i i 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 i on submittal of LERs. i Between October 25 and November 16, 1903, the NRC held five regional workshops to discuss the new LER rule (10 CFR 50 ;3) and the revised emergency notification rule (10 CFR 50.72). Supplement 1 to NUREG-1022 was published in j February 1984 to provide a summary of answers to questions asked during the workshops. Supplement 2 to NUREG-1022, issued in September 1985, contained evaluat'ons 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 (AE00) by EG&G, Inc., at Idaho National Engineering Laboratory. The report identifies deficiencies in LER content and recommends corrective actions. NRC Reaulatory Impact 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 4 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. i Specific industry concerns included the need for reporting inadvertent actuations of engineered safety feature (ESF) equipment 4 e 4 actuation of ESF equipment involving no safety significance e plant shutdowns required by plant technical specifications even though e the action statements of the technical specifications were being met grass fires not affecting plant safety e radiation exposures in excess of regulatory limits e Revision of NUREG-1022 Partially in response to the industry's concerns regarding event reporting described in NUREG-1395, the NRC sponsored four additional regional workshops on event reporting during September to November 1990. 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, Revision 1, to NUREG-1022 is issued to encompass ami supersede NUREG-1022 and Supplements 1 and 2. The intent of Second Draft, A-2 NUREG-1022, Rev. I

this revision is to clarify reporting required by 10 CFR 50.72 and 50.73, as interpreted by the' associated Statements of Consideration, without changing. the reporting requirements of 10 CFR 50.72 and 50.73. i l i I i i i I i i A i l 4 .1 i I i l a i i 4 i I i l j i i i i Second Draft, i A-3 NUREG-1022, Rev. 1 i

i 4 e d i i I APPENDIX B i i EM2RGENCY NOTIFICATION SYSTEM PROCESS I i j i i, 1 i 3 r 1 l s i i i i i I 1 1i i ii 1, i I e i s i 1 1 l 3 i i 1 i, I J t i Second Draft, d 4 NUREG-1022,-Rev. 1 4 i 4 l n--- .:,.,.+- -.-.2

i t NRC Prompt Response Personnel Headquarters Operations Officer i The U.S. Nuclear Regulatory Commission (NRC) Operations Center is continuously staffed with an NRC headquarters operations officer (H00), who holds a degree 4 in engineering and works for the Office for Analysis and Evaluation of Operational Data (AE00). H00s are trained to receive licensee notifications via the emergency notification system (ENS) made under Titel 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 H00 is usually on duty during normal working hours to help with the more frequent communications experienced during the work day. j Each H00 has previous nuclear experience and receives extensive classroom and simulator training on both boiling-water and pressurized-water reactor systems i at the NRC Technical Training Center. j Although H00s have a good general understanding of nuclear power plants, they ) do not have expert knowledge of each specific plant. The H00s ask questions a and rely on the licensees to explain plant-specific details, terms, and the limiting conditions fur oparation of related technical specifications, to ensure they understand the significance of the event and are able to answer pertinent questions. The H00s 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 H00 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 Iristitute 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 H00 notifies the emergency officer (E0). The E0 is assigned on a weekly rotation from NRC staff members of the Senior Executive Service, and is on call 24 hours. These are typically NRR division directors, assistant division directors, or branch chiefs, who are responsible for the NRC response to an event. The E0 decides which other NRC managers should be informed to participate in responding to the event. The E0 also participates in deciding whether the NRC Operations Center and/or the applicable NRC regional incident response center will be partially or fully staffed to continuously monitor the event. Regional Duty Officer Second Draft, B-1 NUREG-1022, Rev. 1

The H00 promptly informs the regional duty officer (RD0) of any ENS notification affecting the RD0's NRC region. The RDO, who is a senior NRC employee (typically a section chief, branch chief, or division director) in the applicable NRC region, is assigned a weekly rotation and is on call 24 hours. The RD0 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 120 can not obtain a clear understanding of an event, the RD0 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 re.gulations 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 NP,C 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 tha NRC regional incident response center may be fully activated and a sits team sent to the plant. Specific NRC staff may monitor the progress of the event from the NRC e Operations Center and/or regional i?cident response center and an NRC team may be sent to the plant. A resident inspector may be requested to immed'ately investigate, monitor, and report back to the NRC region and/or headquarters. Conference calls among NRC headquarters, region, and licensee management may be established. e The E0, RDO, and H00 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. Second Draft, B-2 NUREG-1022, Rev. 1 i l

i 1 l The RD0 may receive the notification and contact the resident inspector e for additional information. Additional NRC Operating Event Review Each working day the NRR Events Assessment Branch (EAB) and the AE00 Reactor Operations Analysis Branch (R0AB) obtain copies of notifications of events that were received in the NRC Operations Center since-the beginning of the 4 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 EAB and ROAB personnel screen the notifications and regional daily reports to identify events that are potentially significant. A 1 telephone conference follows at a preset time in the morning among representatives of EAB, ROAB, NRR's Generic Communications and Vendor Inspection Branches, 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 areventive action is being taken, and whether that action is sufficient. If t1e 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 ROAB personnel to identify similar occurrences and assess generic applicability. The event j is significant if any of the following occurred: potential or actual degradation occurred in safety-related equipment or e structures, fuel integrity, the primary coolant pressure boundary, or containment release of radioactivity (in excess of 10 CFR Part 20 limits) occurred e the plant was operated outside technical specification limits e a scram with complications occurred e other conditions warranted attention by NRC e If the event is classified as significant, senior NRC management are informed at the next weekly events briefing meeting. Briefing information, including 4 event summaries and diagrams, are placed in the Public Document Room (PDR). The event also is entered into the EAB 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 AE00 and available in the NRC PDR). Additional event followup actions performed by NRR, the appropriate NRC regional office, and AE00 personnel may include consulting with the Executive Second Draft, B-3 NUREG-1022, Rev. 1

Director for Operations in the selection of an incident investigation team (!IT), 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. Depending on the number or types of event notifications by licensees, NRR also may issue NRC generic letters, bulletins, and information notices. ] Second Draft, B-4 NUREG-1022, Rev. 1

3 i I l APPENDIX C LICENSEE EVENT REPORT REVIEW PROGRANS' I .1 i; i 1 4 1 l. i i i e 4: i j I 1 1 i i 1 1 i i 1 1 i J 4 i i Second Draft, l NUREG-1022, Rev. I 1 i

j

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 I detail the events or conditions required to be reported, LERs provide information for detailed studies of events or conditions that might affect the l health and safety of the public. l Variations in LER counts from plant to plant can result from numerous factors, i 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. j 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 (AE00) and its contracted national laboratories, screen, classify, categorize, trend, assess, and store the data for each LER. Those events and conditions, both slant-specific and generic, that appear to be important to safety are furtier analyzed or evaluated. From this review process, the NRC determines further actions such as (1) a special study initiated to propose ravisions to regulatory programs, (2) reporting as an abnormal occurrence to Congress, or (3) dissemination to the U. S. nuclear power industry through generic comunications and to the international community through the Nuclear Energy Agency (NEA). The NEA is part of the Organization for Econcmic Cooperation and Development and gathers information from its member countrias 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 e potential severe core damage to discover' emerging trends or patterns of potential safety significance e to identify events that are important to safety and their associated e safety concerns and root causes and to determine the adequacy of corrective actions taken to address the safety concerns to assess the generic applicability of events e 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 i occurred. I An analysis of trends and patterns in operational experience identifies repetitive events and failures and searches past operating history for similar Second Draft, C-1 NUREG-1022, Rev. 1 1 i ?

events and failures to determine if the frequency of such events or failures is significant enough to be a cause for concern. When appropriate, an NRC' bulletin or information notice is issued or a generic study initiated to focus on the nature, cause, con. sequences, 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 becon:e a safety significant factor because of repetition or, more accurately, the i frequency of occurrence. AE00 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 AE0D recommendations for regulatory or industry actions. 1 Special study reports document accelerated assessments of significant 4 e operating events and contain recommendations for remedial actions, if 1 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 e to have little safety significance. 4 i AE0D 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. AE00 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 4 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. AE00 also maintains LER information in the trends and patterns data base at Idaho National Engineering Laboratory (INEL). This data base supports such specific AE00 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 Comission 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 a Second Draft, C-2 NUREG-1022, Rev. 1

-. ~. -. i l i l 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, 4 significant event notifications, and operations and maintenance reminders. l U.S. vendors and nuclear steam system suppliers, as well as other countries and international organizations, use LER data as a source of operational I experience data, i l .i j-j i I i i I 2 1 4 1 1 1 i 1 i l 1 i i l 1 i i 4 4 i I i Second Draft, j C-3 NUREG-1022, Rev. I 4 .~

'l APPENDIX D 10 CFR 50.72 INCLUDING STATEMENT OF CONSIDERATIONS .l Published in the Federal Register On August 29, 1983' (Vol. 48, No. 168, pages 39039-39046) NOTE: This Federal Register notice does not provide a. current version of 10 CFR 50.72, which has been amended severai times since 1983. Its.pur)ose here is to present the Statement of Considerations, which. explains tie basic reporting requirements of 10 CFR 50.72. 1 I Second Draft, NUREG-1022, Rev. 1 q~ a

i 3904d Federal Register / Vol. 48. No. 2G8 / Monday. Aupet 29,1983/ Rules and Regulations Accordingly, this rulemaking includes prepare detsiled written reports for Conditions o/ Licenses (f Sa54) en amendment to 10 CFR 50.54 that certain events (48 FR 33850). A few commenters said that the

" 'dd PP pr ate noti ation

d. Coordination with Licensee's

" Commission already has the ability to ,q n en d Emergencg/an ~ enforce its agulations and does not operating license of each nuclear utilization facility licensed under section The current scheme for licensees. need to incorporate the items as now 103 or 104b. of the Atomic Energy Act of cmergency plans includes four pmposed into conditions oflicense. 1954, as amended. 42 U.S.C. 2133,2134b. Emergency Classes. When the licensee The Cornmission has decided to These facilities generally are the declares one of the four Emergency promulgate the proposed revision of commercial nuclear nower facilities Classes,it must report this to the 15011. "Cenditions of Licenses." in which produce electricity for public Commission as required by 5 50.72. The order to satisfy the intent of Congress as consumption Research and test reactors lowest of the four Emergency Classes. expressed in Section 201 of the Nuclear are not subject to the license condition Notification of Unusual Event, has Regulatory Commission Authorization as they are licensed under section 104a. resulted in unnecessary emergency Act for Fiscal Year 1980.This Act and or 104c. of the Act. Under the declarations. Events that fall within the its relationship to i 50 54 are discussed amendment to 10 CFR 50.54. licensees Unusual Event class have been neither in detailin the Federal Register notice falling under sections 103 or 104b. would emergencies in themselves nor for the proposed rule (46 FR 61894). be required, as a condition of their precursors of more serious events that respective operating licenses, to notify are emergencles. Coordination With Other Reporting Although changes to the dermition of Requirements (Fino/ Rule f 50.72) the NRC Immediately of events specified.the Emergency Classes are not being Seven commenters said that the NRC in 10 CFR 50J2. should coordinate the requirements of to 2 U""#C##5 ry RePorf# repo ing c e oul u timately CFR 50.72 with other rules, with Several categories of reports required climinate " Unusual Event" as an NUREG-nG54. " Criteria for Preparation by i 50.72 are not useful to the NRC. Emergency Class requiring notification Among these categories are reports of: can be adopted consistent with this rule, and Evaluation of Radiological worker injury, small radioactive A proposed rulernaking which would Emergency Response Plans and releases, and minor security problems. redefine the Emergency Classes in Preparedness in Support of Nuclear For example. reports are presently 6 50.47 is in preparation and may soon plants." and with Regulatory Guide 1.16, required if a worker onsite experience be published for public comment.This " Reporting of Operating Information chest pains or amother illness not related final rulemaking makes possible the ..." Many of these letters identified to radiation and is sent to a hospital for elimination of " Unusual Event" as an overlap, doplication. and inconsistency evaluation: or if the vent stack monitor emergency class without further among NRC's reporting requirements. moves upward a few percent yet amendment of l 50J2 by including in he Commission is making a radiation levels remain 100.000 times the category of non Emergencies the concerted effort to ensure consistent below technical specification limits: or if subcategory of"one hour reports." and coordinated reporting requirements, the security computer malfunctions for a

5. Vague or Ambiguous Reporting

%e requirements contained in the ew rninutes. cfiteria revision of 10 CFR 50J2 are being This rulemaking ch,m, ates such coordinated with revision of i 50J3, m reporting requirements from i 5032 and he reporting en.teria in i 50.72 have 5 50.55(e). Appendix E of Part 50, m general clarifies and narrows the been revised in order to clarify their i 20.402. I 73.71, and Part 21. scope of reporting. However, revision of scope and intent.ne criteria were part 73 of the Commission's regulations revised for the proposed rule and in Citing 10 CFR 50.72 as a Bos/s for is necessary to resolve all problems with response to public comment.The Notification (FinalRule f 50.72(a)(d)J security reports. " Analysis of Comments" portion of this Federal Register notice describes 1n A few commenters ob}ected to citing i Terminology. Phrosing. andReportin# more detall specific examples of I 5072 as a basis when making a changes in wording intended to telephone notification.%e letters of Thresholds The various sections of 10 CFR 50 eliminata vagueness or ambiguity. comment questioned the purpose. legal have different phrasing. terminology-11* Anal i' f Co" *

The Commission does not believe that and thresholds in the reporting criteria.

Even when no different meaning is Twenty letters of commerit were it is an unnecessary burden for a intended a change in wording can cause received in response to the Federal licensee to know and identify the basis confusion. Register notice published on December for a telephone notifiestion required by This rulemaking has been carefully 21.1981 (40 FR 61894).' Of the twenty i 50.72.There have been many letters of comment received the vast occasions when a licensee could not tell written to use terminology. phrasing. and reporting thresholds that are either majority its of 20) were from utilities the NRC whether the telephone identical to or similar to those in 5 50J3. wnin8 of operating nuclear powe3 notification was being mada in whenever possible. Other conforming plants. %is Federal Register notice accordance with Technical amendments to Parts 20.21,73, and in described the proposed revision nf to Specifications.10 CFR 50J2, some other i 50.55 and Appendix E of Part 50 are CFR 50J2. " Notification of Significant requirement. or was just a courtesy call. under development. Events, and 10 CFR 50.54. " Conditions Unicss the licensee can identify the of Licenses. A discussion of the more As a parallelactivity to the nature of the report. it is difficult for the preparation of 150.72. on J uly 28.1983, significant corrfments follows: NRC to know what significance the the Commission has published a licensee attaches to the report, and it Licensee Event Report (LER) Rule yb bM[E((.N.$.$c becomes more difficult for the NRC to (i 5013) which requires licensees for Pubuc Demawm Roma. sr:r H 5 tat. Rw., respond quickly and properly to the operating nuclear power plants to Wes e sion.D n aosss event. l

' Fed:rd Register / Vol. 48. N2.108 / Monday. August 29,'1983 / Rules and Replations' 39039 Commission. W'a shington. D.C. 20555; Telephone (301) 492-4973. SUPPLEMEfffAR Y INFORM ATioN:

1. Dackground On February 29,1980. the Commission amended its regulations without prior notice and comment to require timely and accurate licensee reporting of informution following significant events at operating naclear power reactors (45 FR 13434).The purpuse of the rule was i

to proside the Commission with I 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 4 j and timely information to respond to heightened public concern. Although the rule was made immediately effective, comments were solicited. Many commenters 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 j implement Section 201 of the NRC's 1980 FiscalYear Authorizetion Act and change 10 CFR 50.72 to more clearly 1 10 CFR Part 50 specify the significant events requiring licensees to immediately notify NRC. Immediate Notification Requirements The problems and issues which this of Significant Eventa At Operatin9 rulemaking addresses and the solutions Nuclear Power Reactors that it provides can be summarized in five broad areas: AoENCV:Nucicar Regulatory Commission.

1. Authorization ActforFYou ACTION: Final rule, w

Section 201 of the Nuclear Regulatory

SUMMARY

ne Nuclear Regulatory Commission Authorization Act for Commission is amending its regulations Fiscal Year 1900 (Pub. L 96-295) which require timely and accurate provides:

Information from licensees following {s) Section 103 of the Atomic Energy Act of significant events at commercial nuclear 1954 is amended by adding at the end thereof power plants. Experience with existing the following new subsections: f. Each license 4 requirements and public comments on a issued for a utilizat6on facibly under this proposed revision of the rule indicate sedian or section toab shall require as e that the existing regulation should be condition thereof that in case of any accident amended to clstify reporting criteria and which could result in an unplanned release af to require early reports only on those quantities of fission products in excess of matters of value to the exercise of the

I{*$'jhheYm*mIss$on.

censee Commission *s responsibilities.The shall immedistely so notify the Commission. amended regulation will clarify the list Violation of the condition prescribed by this of reportable events and provide the subsection may. in the Commissien's i Commlasion with more useful reports discretion, constitute grounds for ticense regarding the safety of operating nuclear revocation In accordance with section 187 of power plants. this Act.the Commission shat!promptly amend each license for a utilitatten facility EFFECTIVE DATE: January 1.1984. issued under this section or sectica totb. FOR PURTHER $NFORadATION CONTACT: which is in effect on the deu of enactment of Eric W. Weiss. Office ofInspection and this subsection to include the provisions Enforcement. U.S. Nuclear Regulatory required under this subsection. - 1

Fedml Register / Vbl. 48. No.168 / Monday. August 29,19tl3 / Rules and Regulations 39041 /mmediate Shutdown (fino/ Rule airborne concentrations decrease or occurring at operating nuclear pow er f5a72(b)/ffliff until respiratory protection devices are plants. A deadline shorter than one hour Several commenters objected to the utilized.They noted that these events was not adopted because the use of the term. "immediate shutdown." are fairly common and should not be Commission does not want to inte re saying that TechnicarSpecifications do reportable unless the required with the operator a ability to a not use such a term. evacuation affects the entire facility or a an accident or tranwent in the first ew The term is used in some but not all major part of it. cridcal mutu. Technical Specifications. Con sequently. ne Conunission agrees. The wotdinS Thesefore. based on these rormwnts the Commission has revised the f this criterion has been changed to and its experience. the NRC has reporting criterion in question

The final yclude only those events which established a "four. hour report,, as was rule requires a report upon the initiation sigmficantly hamper the ability of site suggested.

of any nuclear power plant shutdown perso np eee {0, s fe pe a io A'8C'0' 8C#8*8 III""I A"I' required by T ecenical Specifications. One commenter was concerned that fM72/bf/2)(ii)/ Plant Operuting cnd.Em a:;ency events occurring on land owned by the Several commenters said that reactor. Procedures (fino/ Rule f Sa72(b)(1//li)) utility adjacent to its plant might be scrams particularly those scrams below Severd commenters said that the reportable.This is not the intent of this power operation, should not require reportmg cnteria should not make reporting requirement.The NRC is notification of the NRC within one hour. concerned with the safely of plant and in response to these comments, the re erence to plant operating and personnel on the utility's site and not Commission had changed the repntting with non-nuclear actMues on land deadline to four hours. However, the o ld ske operators to long to adjacent to the plant. Commission does not regard reactor a decide whether a plant dondition was covered by the procedures. Explicit Threats (Tino/ Rule scrams as "non. events." as stated in b.The procedures cover events that f M 72(b)/f)(vi)) some letters of comment. Information related to reactor scrams has been are not of concern to the NRC. and A few commenters said that the intent

c. He procedures vary from plant to of the term. "explicitla *Mestens." was useful in identifying safety-related plant-unclear. nose come crinh & icted problems.%e Commission agrees that While the plant operating personnel whatlevelof threat m invoo u.The f ur h urs is an appropriate deadline for should be familiar with plant term," explicitly threatm. " Ns been this reporting requirement because these procedures.11 is true that procedures deleted from the final rule. hastead. the events are not as important to vary from plant to plant and cover final rule refers tc "any event that poses immediate safety as are some other events other than those which an actual threat to the safety of the events.

compromise plant safety. However,. he nuclear power plant" [I 5032(b)(1)(vi)] Radioactive Release Threshold (Fino/ t wording of the reporting criteria has and gives examples s'o that it is clear the Rulef272(b)(2)(iv)) .been modified (i 5052(b)(1)(II)in the Commission is interested in real or Several commenters said that the final nile) to narrow the reportable actual threats as opposed to threats threshold of 25% of allowable limits for events to those that significantly without credibility. compromise plant safety, radioactive releases was too low for Notwithstanding the fact that the Notification Timing (Tino/ Rule one-hour reporting. procedures vary from plant to plant, the f272(bg2)) Based upon these comments and its Commission has found that this criterion %e commenters generally had two experience, the Commission has results in notifications indicative of points to make regarding the timing of changed the threshold of reporting to serious events.The narrower. more reports to the NRC. First, the comments those releases exceeding two times Part specific wording will make it possible supported notification of the NRC after 20 concentrations when averaged over a for plant operating personnel to identify appropriate State or local agencies'have period of one hour.nis will eliminate reportable events under their specific been notified. Second, two commenters reports of releases that represent, operatirg procedures. requested a new four-to six-hour report negligible risk to the public. Building Evocuation(fino/ Rule category for events not warranting a The Commission has found that low

E"##lbll#IIN#ll report with one hour.

level radioactive reteases below Iwo Allowing more time for reporting tiraes Part 20 concentrations do not, in Ten commenters said that the some non-Emergency events would theoselves, warrant immediate proposed I 50J2(b)(6)(lii) regarding lessen the impact of reporting on the radio 0gical response. "any accidental. unplanned or individuals responsible for maintaining This 3.aragraph requires the reporting uncontrolled release resulting in the plant in a safe condition.1.!miting of those events that cause en unplanned evacuation of a building" was unclear the extension of the deadline to four or uncontrolled release of a significant and counterproductive in that it could hours ensures that the report is made amount of radioactive materi' i to'offsite a cause reluctance to evacuate a bolding. when the information is fresh in the areas. Unplanned relesses should occur Many of these commenters stated that minds of those involved and that it is infrequently; however, when they occur, the reporting ofin-plant releases of morelikely to be made by those at least moderate defects have occurresi rsdioactivity that require evacuation of involved rather than by others on a later in the safety design or operational individual rooms was inconsistent with shift. control established to avoid their the general thrust of the rule to require Other, more significant non-occurrence and, therefore, these events reporting of significant events. %ey Emergency evente and all declarations should be reported. noted that minor spills, small gaseous of an Emergency must continue to be PersonnelRod>.oactive Contamination waste releases, or the disturbance of reported within one hour. The one-hour contaminated particulate matter (e.g deadline is necessary if the Commission (Fino1Rulef272(bg2)(v)) dust) may all require the temporary is to fulfillits responsibilities during and Several commentars objected to the evacuation of individual rooms until the following the most serious events use of vsgue terms such as " extensive 3 .. _ ~

39042 Ftdual Regist:r / Vol. 48. No.1Ga / Mond:y. August 29. 1983 / Ruics and Regulations onsite contamination" and "readily from a license condition or technical the containment). Examples of this type removed"in one of the reporting criteria specification. of situation include: of the proposed rule.' Pomgroph 272(b)(1)//i), (a) Fuel cladding failures in the Based on this comment. new criteria encompsasing esents previously reactor, or in the storage pool. that have been prepared that use more classified as Unusual Events and some exceed expected values, or that are specific terms. For example. one new events captured by proposed unique or widespre.id. or that are criterion requires reporting of "Any ( 50.7 (b)(1) was added to provide for caused by unexpected factors, and event requiring the transport of a consistent, coordinated reporting would involve a release of significant rcdioactively contaminated person to an requirements between this rule end to quantities of fission products. offsite medical facility for treatment." CFR 50.73 which has a similar provision. (b) Cracks and breaks in ihn piping or Experience with telephone notifications Public comment suggested that there reactor vessel (steel or prestressed made to the NRC Operations Center should be similarity of terminology. concretc) or major con ponents in the suggeststi et this new criterion will be phrasing and reporting thresholds primary coolant circuit that have safety easily M.derstood. between j 50.72 and i 50.73.The intent relevance (steam generators. reactor 'of this paragraph is to capture those coolant pumps. valves, etc.). III. Paragrep, -by-Paragraph Explanation events where the plant including its (c) Significant welding or material n cf the Rule principal safety barriers, was seriously defects in the primary coolant system. Pomgmph 50.72(of reflects some degraded or in an unanalyzed condition. (d) Serious temperature or pressure consolidation oflanguage that was For example, small volds in systems transients. repeated in various subparagraphs of designed to remove heat from the (e) Loss of relief and/or safety valve the proposed rule. In general, the intent reactor core which have been previously functions during operation. and scope of this paragraph do not shown through analysis not to be safety (f) Loss of containment function or reflect any change from the proposed significant need not be reported. integrity including: ruta. liowever, the accumulation of voids that (i) Containment leakage rates Several titles were added to this and could inhibit the ability to adequately exceeding the authorized limits, subsequent sections.For example, remove heat from the reactor core, fil) Loss of containment isolation p:ragraph 50.72(b)is titled "Non. particularly under natural circulation - valve function during tests or operation. Emergency Events" and it has two conditions. would constitute an ~ (iii) Loss of main steam isolation subparagraphs: (b)(1), titled. "One-Hour unanalyzed condition and would be valve function during test or operation. RIports" and (b)(2)."Four-Hour reportable. In addition. voiding in instrument lines that results in an R ports."The events which have a one, W W d miW WW h:ur deadline are those having the erroneous indication causing the potential to escalate to an Emergency operator to misunderstand the true ca{abillt lant is also an 58 g condition of thefition and should be CI:ss The four-hour deadline is enco as ing a portion of p posed explained in the analysis of paragraph ana ed con g, g;e ,,ngi e nt 3(a (11$). ng the r q irements I0U ' 8 s th Broph Sa72(b)(1)(illa) requires a reporting of "The initiation of any of to CFR 50.72 and 50.73 similar in nuc ear plant shutdown required by and experience to determine whether an Technical Specifications. Although the unanalyzed condition existed. It is not language increases the clarity of these intended that this paragra h apply to rules and minimizes confusion. intInt and scope have not changed, the minorvariationsinindivi ual The paragraph has also been change in wording between the parameters. or to problems concerning reworded to make it clear that it applies pro sed and final rule is intended to a y that prompt notification is single pieces of equipment.For example, only to acts of nature (e.g tornadoes) at any time, one or more safety-related and external hazards (e.g., railroad tank required once a shutdown is init!ated. components may be out of service due car explosion). References to acts of In response to public comment. the to testing maintenance, or a fault that sabotage have been removed, since term "immediate shutdown that was has not yet been repaired. Any trivial these are covered by i 73.71. In addition, usId in the proposed rule is not used in single failure or minor error in threats to personnel from intemal th2 final rule.The term was vague and performing surveillance tests could hazards (e.g., radioactivity releases) that unfamiliar to those licensees who did produce a situation in which two or hamper personnel in the performance of n:t have Technical Specifications usin8 more often unrelated, safety. grade necessary du!ies are now covered by th2 term. components are out-of service. Paragraph 50.72(b)(1)(vi). nis paragraph This reporting requirement is intended Technically, this is an unanalyzed covers those events inniving an actual 13 capture those events for which condition.However, these events should threat to the plant frem an external Technical $pecifications require the be reported only if they involve condition or natural phenomenon, and initiation of reactor shutdown.nis will functionally related components or if where the threat or,amage challenges l provide the NRC with early warning of they significantly compromise plant the ability of the plant to continue to safety significant conditions serious safety. When applying engineering operate in a safe manner (including the enough to warrant shutdown of the judgement, and there is a doubt orderly shutdown and maintenance of plant. regarding whether to report or not, the shutdown conditions).ne licensee Paragraph San (b)(I)(i)(B) was added Commission's policy is that licenseca should decide if a phenomenon or 13 be consistent with existing should make the report. condition actually threatens the plant. requirements in I $0.54(x) and the Finally, this paragraph also Includes For example. a minor brush fire in a existing I 50.72(c) as published in the material (e.g metallurgical or chemical) remote area of the site that is quickly Federal Register on April t.1983 (48 FR problems that cause abnormal controlled by fire fighting personnel and. 13966) which require the licensee to degradation of the principal safety as a result, did not present p Arcat to notify the NRC Operations Center by barriers (i.e the fuel cladding, reactor the plant should not be reported. telephone when the licensee departs coolant system pressure boundary, or However, a major forest fire. larg.e-scale

Fed:ral Regi:ter / Vol. 48. No.168 / Monday. August 29. 1983 / llules and Rcgiilati:ns 39043 j flood, or major earthquake that presents

5. Plant monitors necessary for is punibic. because these personnel will j

e clear threst to the plant should be-accident assessment. have a better knowledge of the reported. As another example, an . Pomymph Sa72(b)(1)(vi). circumstances associated with the vent. Industrial or transportation accident encompassing some portions of the Reports made within four hours of the which occurs near the site. creating a proposed il 50.72[h) (2) and M). has event should make this possible while plant safety concem, should be been revised to add the phran, not imposing the more rigid one hour reported. " including fires, toxic ges Maases, or requireim nts. Foraymph M.72(b)(1)(iv), rudwactis e releascs."".1.'.s addition The typorting requirement in encumpassing events previously covers tie "es acuatic. ' portion of pumgr.:ph.W 72/b//2)(i)in similar tu a classified as Unusual Events, requires puragraph 50.7:tulp)(iii) of the proposed regimement in i 50.73. Moreover cxtept the reporting of those events that resuti rule. This change ia wording for the final for referring to a shutdown reactor. this in either automatic or manual actuation rule was made in response to public reporting requirement is also similar to of the ECCS or would have resulted ir, cornments discussed above. the "One.Ilour Report"in activation of the ECCS if some While paragraph co.72(b)(1)(iii) of the g m;2(bl(1)(ii). However this parngrnph component had not failed or an operator final rule primarily captures acts nf applies to a n: actor in shutdown action had not been taken. nature paragraph 50.72(b)(1)(vi) condition. Events within this For example.if a valid ECCS signal captures other events, particularly acts requirement have less urrency and can were generated by plant conditions and by personnel.The Commission believes be reported within four hours as a "Non-the operator were to put all ECCS this arrangement of the reporting criteria Emergency." pumps in pull.to. lock, though no ECCS in the final rule lends itself to more Pomproph 272(b)/2)(ii)(proposed discharEe occurred, the event would be precise interpretion and is consistent 50.72(b)(5))is made a "Non. Emergency" reportable. with those pubic comments that in response to public comment, because A " valid signal" refers to the actual requested closer coordination between the Commission agrees that the covered plant conditions or parameters the reporting requirements in this rule events generally have slightly less satisfying the requimments for ECCS and other portions of the Commission's urgency and safety significance than initiation. Excluded from this reporting regulations... those events included in the "One.Ilour requirement would be those instances This provision requires reporting of Reports " where instivment drift. spurious signals. events, particularly those caused by acts %e intent and scope of this reporting human error, or other mvalid signa'.s of personnel, which endanger the safety requirement have not changed from the caused actuation of the ECCS. lloweser, of the plant or interfere with personnel proposed rule. This paragraph is such events may be reportable under m performance of duties necessary for intended to capture events during which other sections of the Commission a safe plant operations. an ESF actuates either manually or regulations based upon other details:in The licensee must exercise some automatically, or falls to actuate. ESFs particular, paragraph 50.72(b)(2){li), judgment in reporting under this section. requires a report within four hours if an For example, a small fire on site that did are provided Io mitigate the Engineered Safety Feature (ESF)is not endanger any plant equipment and consequences of the event: therefore.(d 1) the) should work pmperly when calle actuated. that did not and could not reasonably be Experience with notifications made expected to endanger the plant, is not upon and (2) they should not be pursuant to i 50.72 has shown that reportable. challenged unnecessarily.The, Commission in interested bnth m events events involving ECCS discharge to the Pomgmph 50.72(b)(1)of the pmposed vessel are generally more serious than rule was split into f 50.72(b)(11/ii) and where an ESF wus needed to mitigate ESF actuations without discharge to the f 50.72(b)/2/(i)in the final rule in order the cnnsequences of the event (whether vessel. Based on this experience, the to permit some type of reports to be or rmt the equipment performed Corr missinn has made this reporting made within four hours instead of one properly) and events where an ESF cnterion a "One. Hour Report." hour because these reports have less openited unnecessarily. Pomgmph 50.72(b1(11/v). safety significance. In terms of their "Actuntion" of multichannel ESF encompassing events previously combined effect the overallintent and Actuation Systems is defined as classified as Unusual Events covers scope of these paragraphs have not actuation of enough channels to those events that would impair a changed from those in the proposed rule. complete the minimum actuation log,c. i licensee's ability to deal with an Since the types of events intended to be %crefore single (hannel actuations, accident or emergency. Notifying the captured by this reporting requirement whether caused by failures or otherwise.. NRC of these events may permit the are similar to i 50.72(b)(1)(ii). except are not reportable if they do not NRC to take some compensating that the reactor is shut down, the reader complete the minimum actuation logic. measures and to more completely assess should refer to & :.planation of Operation of an ESF as part of a the consequences of such a loss should i 50.72(b)(1)(ii' for e tore details on planned test or operational it occu.r.during an acciderit or intent. evolution need not be reported. cmergency. Pomgmph 50.72(b)(2/ Although the flowever,if during the test or Examples of events that this criterion reporting criteria contained in the evolution the ESF actuates in a way that is intended to cover are those in which subparagraphs of I 50.72(b)(2) were in is not part of the planned procedure, any of the following are not available: the proposed rule, in response to public that actuatinn should be reported.For

1. Safety parameter display system comment the Commission established example.if the normalreactor shutdown -

(SPDS). this "Non. Emergency" category for procedure requires that the control rods

2. Emerscocy Response Facilities those events with slightly less urgency be inserted by a manual reactor trip, the (ERF's).

and less safety sMnificance that may be reactor trip need not be reported.

3. Emergency communications reported within four hours instead of However,if conditions develop during facilities and equipment including the one hour.

the shutdown that require an automatic Emergency Notification system (ENS). The Commission wants to obtain such reactor trip, such a reactor trip should

4. Public prompt Notification System' reports from personnel who were on be reported.The fact that the safety meluding sirens.

shift at the time of the evert. when this

9044 Federd Regist:r / Vol. 48. No.168 / Mondry. August 29, 1983 / Rulis cnd R:gultti:ns' nelysis assumes that an ESF will service to perform maintenance, and the criterion. For example, the Commission ctuate automatically during an event Technical Specifications permit the is increasingly concerned about the ces not eliminate the need to report resulting configuration, and the system - effect of a loss or degradation of what ist actuation. Actuations that need not or component is relumed to service had been assumed to be nonessential i e reported are those initiated for within the time limit specified in the inputs to safety systems.Therefore, this casons other then to mitigate the Technical Specifications, the action paragraph also includes those cases onsequences of an event (e.g at the need not be reported under this where a service (e.g., heating. iscretion of the licensee as part of a paragraph.However.if while the ' ventilation, and cooling) or input (e.g, lanned procedure). component is out of service. the licensee compressed air) which is necessny ter identifies a condition that could have reliable or long term operatu.n of a l Porograph 50.72(b)(2)(iii)(proposed prevented the system frora perforrning safety system is lost or chgraded. Such i C.72(b)(4)) has been revised and its intended function (e.g., the licensee loss or degradation is 'eportable,if the .implified. finds a set of relays that is wired proper fulfillment of ue safety function The words "any instance of personal incorrectly), that condition must be is not or can not be assured. Failures

rror, equipment failure, or discovery of reported.

that affect inputs er services to systems i: sign or procedural inadequacies" that It should be noted that there are a that have na safety function need not be +peared in the proposed rule have been limited number of single-train systems reported. eplaced by the words " event or that perform safety functions (e.g the Finally, the Commission recognizes

cndition." His simplification in.

High Pressure Coolant Injection System that the licensee has to decide when anguage is intended to clarify what was in BWRs). For such systems. loss of the personnel actions could have prevented a confusing phrase to many of those single train would prevent the.. fulfillment of a safety function.For who commented on the proposed rule, fulfillment of the safety im.ction of that exa mple, when an individual improperly 'tiso in response to public comment this system and, therefore. must be reported operates or maintains a component, that vporting requirement is a "Non-even though the plant Technical person might conceivably have made Emergency" to be reported within four Specifications may allow such a the same error for all of the functionally sours instead of within one hour. condition to exist for a specified length redundant components (e.g,if an His paragraph is based on the of time. Also,if a potentially serious issumption that safety-related systems human error is made that could have individual incorrectly calibrates one bistable amplifier in the Reactor and structures'are intended to mitigate prevented fulfillment of a safety the consequences of an accident. While function, but recovery factors resulted in Protection System, that person could paragraph 50.72(b)(2)(ii) applies to the error being corrected. the error is conceivably incorrectly calibrate all actual demands for actuation of an ESF. still reportable. bistable amplifiers). However, for an paragraph 50.72(b)(2)(ill) covers an ne Commission recognizca that the event to be reportable it is necessary event where a safety system could have application of this and other paragraphs that the actions actually affect or fziled to perform its intended functidn of this section involves a technical involve components in more than one because of one or more personnel errors, judgment bylicensees. In this case, a train or channel of a safety system, and mcluding procedure violations: techmcal judgment must be made the result of the actions must be i ipment failures; or design, anal sis, whether a failure or operator action that undersirable from the perspective of I f: neation. construction, or proce ural disabled one train of a safety system, protecting the health and safety of the ) d:ficiencies.The event should be could have, but did not, affect a - public.The components can e I reported regardless of the situation or redundant train. If so, this would functionally redundant (e.g., two pumps ccndition that caused the structure or constitute an event that "could have in different trains) or not functionally system to be unavailable. prevented" the fulfillment of a safety redundant (e.g, the operator correctly his reporting requirement is similar function, and, accordingly, must be stops a pumpinTrain A and.Instead t2 one contained in i 50.73, thus reported.- of shutting the pump discharge valve in r:flecting public comment identifying If a component falls by an apparently Train "A." he mistakenly shuts the ths need for closer coordination of random mechanism,it may or may not pump discharge valve in Train "B ). reporting requirements between i 50.72 be reportable if the functionally Pamgmphs 50.72(b)f2)(iv/ (proposed cnd i 50.73. redundant component could fall by the 50.72(b)(6)) has been changed to clarify his paragraph includes those safety same mechanism. To be reportable, it is the requirement to report releases of systems designed to mitigate the necessary that the failure constitute a radioactive material. ne paragraph is i consequences of an accident (e.g ' condition where there is reasonable similar to i 20.403 but places a lower cintainment isolation, emergency doubt that the functionally redundant threshold for reporting events at filtration). Hence, minor operational train or channel would remain commercial power reactors. The lower evInts such as valve packing leaks, operational until it completed its safety threshold is based on the significance of which could be considered a lack of function or is repaired. For example, if a the breakdown of the licensee's program control of radioactive material, should pump fails because ofimproper necessary to have a release of this size, n t be reported under this paragraph. lubrication, there is a reasonable rather than on the significance of the System leaks or other similar events expectation that the functionally irnpact of the actual release The may. however, be reportable under other redundant pump, which was also existing licensee radioactive material piragraphs. Improperly lubricated, would have also effluent release monitoring programs This paragraph does not include those failed before it completed its safety and their associated assessment cases where a system or component is function. then the failure is reportable capabilities are sufficient to satisfy the l removed from service as part of a and the potential failure of the intent of 50.72(b)(2)(iv). l planned evolution. in accortlance with functionally redundant pump mubt be Based upon public comment and a l cn appioved procedure, and in. reported. reevalustMy the Commission staff, accordance with the plant's Technical Interaction between systeins. the repo' areshold hee been Specifications.For example,if the particularly a safety system and s non. changed.. m "25%" in the proposed rule licensee removes part of a system from safety system. is also included in *his to "2 times"in the final rule and has

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. -Second Draft, NUREG-1022, Rev. 1

Fedtral Regist2r / Vol 48. No.108 / Monday. August 29, 1983 / Rules and Regulatiims 39045_ been reclassified as a "Non.Emergeroy" respond because of media or public t.ist of Subjects in to CFR Part 50 to be reported within four hours inswad attention. An rust. Classific'd information. Fire of within 1 hour: Pomgruph.M 72/c/ (proposed 50.72(c)) prevenuon. Incorpnrution by reference. Also this reporting requirement has has remained essentially unchanged I"I "'8" *" *'"' dI Td " 'I"" been changed to make a more uniform from the proposed rule. except for p wer plants and reactors. Penalty, requirement by referring to specific addition of the title " Followup U"d"" P"*Cd"" D"'# #" A release criteria instead of referring only Notification" and some renumbering. crueria. Reporting unit re cordkeeping to Technical Specifications that may This paragraph is intended to provide vary somewhat among facilities. the NRC with timely notification when P"n nu nts. Puranant to the Atomic Fncrgy Act of This report ng requirement is intended an event becomes more serious or to capture those events that may lead to additionalinformation or new analyses 1954, as amended, the Energy .in accident situation where significant clarify en event. Reorganization Act of 1974. as amended, amounts of radioactive matenal could This paragraph also permits the NRC and section 55* and 553 of Title 5 of the be released from the fucility. Unplanned to maintain a continuous United States Code, the following releases should occur infrequently: communications channel because of the amendments to Title 10. Chapter L Code however. if they occur at the levels need for continuing follow up of Federal Regulations, Part 50 tre specified, at least moderate defects have information or because of published as n' document subje.:t to relec mmunicati ns problems. codification. pe a onal on I stabl ed to avoid their occurrence and, therefore, such IV. Regulatory Analysis PART 50-DOMESTIC UC2NSING OF events should be reported. ne Comm.ission has prepared a "RODUCTIOt4 AND UT1ULATION - Normal operating limits for regulatory analysis on this regulation. FACIUTIES radioactive effluent releases are based The analysis examines the costs and on the limits of to CFR Part 20 which benefits of the Rule as considered by the

1. The authority citation for Part 50 establishes maximum annual average s

co o e g continues to read as follows: concentration,in unrestricted areas. This g reporting requirement addresses copying for a fee at the NRC Public Authority Secs.103.104.161.1a2.1a3,1n6. oncentrations averaged over a one Document Room.1717 H Street. NW., 189 es Stat. saa. s37,94a. ess. 954. 955. 950. as amended sec. 234, as Stai 1244. as amended our per od and represents less than Washington, D.C. Single copies of the radioac ive ma riaI perrnitted Io be analysis muy be obtained from Eric W. [42 US.C 2133,2134,2201. 2232. 2233,2236. 2239. 22a2h seca. 201. 202. 2m se Stat.1242. Weiss. Office of Inspection and. l 2 as a 42 C 5641.5a42. Po graph 72(b)(2 (v (proposed Enforcement. U.S. Nuclear Regulatory y4 rule 50.72(b)(7)} has three changes. He Commission. Washington. D.C. 20555. de N first eliminates the phrase " occurring Telephone (301) 4924973. eat, sec.10. 92 Stat. 2951142 US.C 5851). onsite" because it is implied by the V. Paperwork Reduction Act Statement Sections m58. 291 and 50.92 also issu5 scope of the rule.%e second replaces under l'uh. L 97-415. se Stat. 2073 (42 U.S.C " injury involving radiation" with i sasmd eder on. The information collection " radioactively contaminated person /- requirements contained in this final rule 122. ca Stat. 939 (42 US.C 2152). Sections His change was made because of the have been approved by the Office of som. mat minu issued under sec.184.se Stat. s difficulty in defining injury due to Management and Budget pursuant to the 954. as amended (42 U.S.C 2234). Sections radiation. und more importantly. Paperwork Reduction Act. Pub. L 96-511 50.1tn.50.102 also issued under sec. tu os becacfe 1(,CFR Part 20 captures events (clearance number 3150-0011). Stat. 9ss t42 US.C. 2236). VI. Regulatory Flexibility Certification For the purtmses of sec.223.sa Stat.958 as involvmg radiation exposure. nmended (42 tt.S.C 2273), il M10 ta). (bl. %e third change. m response to public comraent was to make this .in accordance ivilh the Regulatory and (cl. 50 44. m40. 50.4a 50.54. and so.no(a) reporting requirement a four. hour Flexibility Act of 1980. 5 U.S.C. 005(b). are issued under sec.101b. se Stat. 94a, as notification.instead of one hour the Commission hereby certifies that amended 142 U.S.C 22nt(b)h Il mio (b) and riolification.This change was made this regulation will not have a (c) and m54 are issued under sec. teil. es because these events have slightly less sigmficant economic impact on a Stat.949. un amended (42 US.C 22mti)); and safety significance than those required , substantial number of small entities. Il ms(c) 25Hbl. 50.70. m71. 50.72, and to be reported witidn one hour, %ls final rule affects electric utilities m7a ard issued under sec.101o. sa Stat. 9so. Paragtnph 50.72(bf/2)(vi)(ndt in thet are dominantin their respective es amended (42 US C. 2201to)). proposed rule) besides covering some service areas and that own and operate events such as release of radioactively nuclear utilization facilities licensed

. A new paragaph (2)is added to contaminated tools or equipment to the under sections 103 and 104b. of the i n54 to rend as follows-public that may warrent NRC attention.

Atomic Energy Act of1954, as amended. also covers those events that would not %e amendments clarify and modify { 5J.54 Conditions of 16 censes, otherwise warrant NRC attention except presently existing notification for the interest'of the news media, other requirements. Accordingly, there is no government agencies, or the public. In new. significant economic impact on (z) Each licensee with a utilization terms of its effect on licensees. this is these licensees, nor do the affected facility licensed pursuant to sections 103 not a new reporting requirement licensees fall within the scope of the or104b.of the Act shallimmediately - because the threshold for reporting definition of"sniall entities" set forth in riotify the NRC Operations Center of the injuries and radioactive release was the Regulatory Flexibility Act or within occurrence uf any event specified in much lower under the proposed ruir. the Small Business Size Standards set i 50.72 of this part. This criterion will capture thos., events forti a regulations issued by the Small previously reported under other criteria Business Administration at 11CFR Part

3. Section 50.72 is revised to read as follows:

when such events require the NRC to 121.

393-16 Federal Regist:r / Vol. 48. No., 'l donday, dugust 29, 1983 / Ruiss end Regulations HD.72 Immediate nouncation (C)In a cmdition not covered by the when averaged over a time period of requirements for operating nuclear po**r. 'srctors. plant's operating and emergency one hour. procedures (B) Any liquid e(fluent release that (a) Cenero/ Reqirements. ' (1),Each (iii) Ara.iatural phenomenon or other exceeds 2 times the limiting combine.i nuclear power re:.ctor under i 50.21(b) externr] condition that poses an actual Maximum Permissible Concentratiri or i 50.22 of this part shall notify the threat o the safety of the nuclear NRC Operatic,ns Center via the power plant or significantly hampers (MPC) (see Note 1 of Appendix B I, Part Emergency Notification System of: site personnelin the performance of 20 of this chapter) at the point of enry (i) The declaration of any of the duties necessary for the safe operation into the receiving water (i.e Emergency Classes spec'fied in the of the plant. unrcstricted areal for all radionuclides licensee's approved Errergency Plan:8cr [iv) Any event that results or should except tritium and dissolved noble (ii) Of those non-Euentency events have resulted in Emergency Core gases, when averaged over a time period specified in paragraph (b) of ths section. Cooling System (ECCS) di charge into of one hour. (Immediate notifications (2)If the Emergency Notification the reactor e olant system as a resu make the, inoperative, th hcensee shall ajiAnY event that results in a major, and (b)(2) of I 20.403 of Part 20 of this System is a required notifi:ations via commerical telephone service, other loss of emergency assessment ch' apter.) dedicated telephe= sveern. or any capability offsite response capability, or (v) A.iy event requiring the transport other method which will ensure that a commumece'n capability (e g* of a r dioactively contaminated person report is made as soon as practical to significant portion of control rnom the NRC Operations Center.: indication. Emergency Notificatio.' te an offsite medical facility for treatment. (3) The licensee shall notify the NRC Sys e roffsite otification e ). immediately after notification of the appropriate State or local agencies and threat to the safety of the nuclear the health and safety of the public or not later than one hour after the time the powerplant or significantly hampers site onsite personnel, or pro'ection of the licensee declares one of the Emergency personnelin the performance of duties environment. for which a news release Q:nes. necessary for the safe operation of the is planned or notifics.fon to other (4) When e 4Fing a report under nulceer powerplant inchig 3res, toxic ' government agencies has been or will be piragraph (a)(3) cf th.s section, the gas releases. or radioactive rebases. made.Such an event may include an licensee shallidemL8 ; (2) Four HourReports. lf not repdrted onsite fatality or inadvertent release of y (i) ne Emergency Clan declared: or under paragraphs (a) or (b)(1) of this radioactively contaminated materials. (11) Either paragraph (b)(1), "One-Hour section. the licensee shall obtify the Report." or paragraph (b)(2). ' Tour. Hour NRC as soon as practical and in all (c)FollowupNotification. With cases, within four hours of the respect to the telephone notifications R: port." as the paragraph of this section requiring notification of the N'n-occurrence of any of the following: made under paragraphs (a) and (b) of (i) Any event. found while the reactur this section. in addition to making the Emergency Event. Is shutdown. that, had it been found required initial notification, each (b) Non-Emergency Events. (1) One-while the reactor was in operation, licensee. shall during the course of the Hour Reports. If not reported as a would have resulted in the nuclear event: helaration of an Emergency Class powerplant, including its pdncipal (1) Immediately report: (i) any further ander paragraph (a) of this section. *he icensee shall notify the NRC as soon as safety barriers, being seriously degraded 3r:ctical and in all cases within one or being in an unanalyzed condition that degradation in the level of safety of the ,to of the occurrence of any of the significantly compromises plant safety. plant or other worsening plant (ii) Any event or condition that resulta conditions, including those that require o owing-in manual or automatic actuation of an., the declaration of any of 11 e &nergency (i)(A) he initiat,on of any nuclear >lant shutdown required by the plant e - Engineered Safety Feature (ESF)* Classes,if such a declar::t!u'n has not i including the Reactor Protection System mp ash made, or W av change rechnical Specifications. (RPS). However, actuation of an ESFJ imm ne Emergmcy dan to mh. or (B) Any deviatinn from the plant's including the RPS. that results from and (111) a termination of the Emergency Technical Specifications authorized is part of the preplanned sequence Ca n. su suant to i 50.54(x) of this past during testing or reactor operation need (2)Immedlofely reportr (i) the results (11) Any event or condition during not be reported. of ensuing evaluations or assessments of >peration that results in the condition of (iii) Any event or condition that alone plant conditions. (li) the effectivenen of he nuclear powerplant including its could have prevented the fulfillment of response or protective measures taken. irincipal safety barriers, being seriously the safety function of structures or and (iii)'.nformation related to plant iegraded; orresults in the nuclear systems that are needed to: behav!or that is not understood. cwerplant being-(A) Shut down the reactor and (3) N aintain an open. continuous (A)In a unanalyzed condition that maintain it in a safe shutdown commun.ication channel with the NRC ignificantly compromises plant safety: condition. (B)In a condition that is outside the (B) Remove residual heat. Operations Center upon request by ~the NRC. esign basis of the plant: or ) atrol the release of radioactive 'othat mqu6mments see innen.de nooscanes et (D) Mitigate the consequences of an August, toes. w NR C by licensed oper= Gas nuclear power accident. For the Nuclear Regulatory Conunission. eacsere era contained etw=hm in tNe ctwear. la (iv)(A) Any airborne radioactive esecular. I samt i no eas, s sax and a n n. release that exceeds 2 times the 8""N" * "***"*d applicable concentrations of the limits b##F#M M88/88-er -. a.; esiepheme men t er of de emc specified in Appendix B. Table II of Part I" h " """ N *** * =al Penuens Center is (3Rlsst.oua 20 of this chapter in unrestricted areas, su m ocaosresoe w f \\

,%deral R'eg'ister / Vhl' 48. Ns.144" / Tuisday, jul'y 6'I'983 [R't$les dd R'c$17.tI5Es ~ 3385'O 2 Acnom Final rule. IL Rulemabg initiatloa SUH8AARY:ne Commission Is amending The Nuclear Ptant Rellability Data its regulations to require the reporting of (NPRL) system is a voluntary program for th* Sporting of reliability data by operational experience at nuclear power nuclea p @ ant Heensus.On plants by establishing the Licensee January 30.1980 (45 FR 6793).' the NRC Event Report (LER) system.The final Published an Advance Notice of rule is needed to codify the LER Proposed Rulemaki.4 tha.t described 'the reporting requirements in crder to NPRD system and tr.vued pui;lle establish a single set of reqilrements comment on an NRC plan to make it that apply to all operating etclear. mandatory. Forty.fourletters were power plants. The final rvie e pplies only received in response to the advanced. to licensees of commercial auclear nodce.nue c mments genuaHy power plants.The final rule will change Pposed making the NFRD system the requirements that define the events mandatory on the grounds that reporting and situations that must be reported, of reliability datr oneuld notbe made a and willdefine the information that "8"I'#S" ""', "'" must be p.mvided in each report. - In Dembr 201 the Commission. amcTava onTc }anuary 1.1984.The decide! tut the requirements for incorporation by reLuence of certain repoziing of operatioani experience data publications listed in the regulations is needed major revision cnd approv'ed the approved by the Directar of the Federal development of anIntegrated - Register as of January 1.1984. - OperationalExperience Reporting ron rumEn INFOR 4AYlON COKrAcn (IOER) system.ne.IOERrysum would FnderickJ.Hebdon Chief. Program have combined, mod &d. and made Technology Branch. Office for Analysis mandatory the existlag Licensee Event and Evaluation of Operational Data. Report (LER) system end the NPRD U.S. Nuclear Regulatory Commission, systen SECY 80-5078 discusses the Washington. D.C. 20555: Telephone (301) IOER systen 492-4480. As a result of the Commission's approvalof the concept of anIOER suretas.NTAny mrons4AMOR system. the NRC published another L Background advance notice on Ja.m :y 15.1981(46 FR 3541).This advance notice ex lained On May 6.1982 the NRC published in w e Dee perahna the FeogralRegister(47 FR19543)* a experience data and described the Notice of Proposed Rulemahg that deficiencies in the existing LER and would m >dify and codify the existing NPRD systems. , Licensee a:,ve.nt Report (LER) systen On June 8.1981, the Institute of Interested persans were invited to submit written e.omments to the Nuclear Power Operations (INPO) announced that because ofits role as an Secretary of the, Commission by July 6 active user of NPRDs data it would - 1962. Numerous comments were assume responsibility for management received..After consideration of the and funding of the NPRD systen comments and other factors involved, Further.1NPO decided to develop the Commisalon has amendad the criteria that would be usedinits proposed requirements published for "*""8'#'ent audits of member utilities public comment by clarifying the scope tc assess the uaequacy of participation and content of the requirements. .m de WRD systen particularly the criteria that define De two precipal deficiencies that. which operational events must be reported. had previously tnde the NPRD systern an inadequate source d reliability data The majority of the comments on the wen 6e inaWy of Hs unittee proposed rule:(1) Questioned the management structure to provide the, meaning and intent of the criteria that necessary technical direction and a low defined the events which must be. level of participation by the utilities. He reported. (2) questioned the need for commitments and actions byINPO reporting certain specific types of provided a basis for confidence that" events, and (3) questioned the need for 6ese two deficiendes would be NUCLEAri REGULATORY certaininfonnation that wouldbe COMMISSION required to be included in an LER. man g ani g of Se tion IH o a statice discusses the withinIMO should overcome the to CFR Parts 20 and 50 comments'th ore detail. previous difficulties associated with Licensee Event Report System , copw of the doa ments are evaltable for pubtle $"O"I fro-evu Ee t

8'"

inspection and copyins for a fee at the Putec AoEncy: Nuclear Regulato 7 Docurnent no es at w H street NW. Washinston, organizations.Further, with INPO , Coramlssion. On focusing upon a utility's participatien in

Federtl Redstt / Vol. 48, No.144 / Tussdty. July 20, 1983 / Rula and Rtgulations 33851 i NPRDS as a specific evaluation comprehensive integrated analytically.

2. Four co'mmenters felt that the lehet parameter during routine management versatile system.

of effort would be increased but not and plant audit activities, the level of %e Brookhaven Study, published as* significantly. utility participation, and the~refore, the BNL/NUREG 51609, fMREQ/CI} 3200,-

3. One commenter felt that the quality and quantity of NPRDS data, riiscusses data collection and storage 4 proposed rule would have a minimal should significantly increase. How aver.

procedures to support multivariate, effect on the level of effort required. j tbs Commission will continue to have an mudcase analysis. While the range of 4.Two commenters felt that the active role in NPRDS by participating in reactor configurations in the U.S. proposed rule would significantly reduce an NPRDS User's Group, by periodically nutlearindustry presents some the number of LERs filed. assessing the quality and quantity of methodological and laterpretative

5. Thirteen commenhw " sed the information available from NPRDS, and problems, these difficulties should not objective of improving IE s.rorting but by auditing the timely availability of the be insurmountable.The Commission felt that changes in the proposed rule information to the NRC.

belleves that the NRC should have as a were needed.These commenters did not Since there was a likelihood that specific objective the development, directly address the resource issue. NPRDS under INPO direction would demonstration, and implementation of

6. Five commenters endorsed the meet the NRC's need for reliability data, an integrated system for collecting and proposed rule and/or felt that it was a it was no longer necessary to proceed analyzing operational data that will significant improvement over the with the IOERS. Hence, the collection of employ the predictive and analytical existing reporting requirements, det:lled technical. descriptions of potential of multicase, multivariate Based on these commgnts and its own significant events could be addressed in analyses. Accordingly, the staff has assessment of the impact of this rule, the a arparate rulemaking to modify and been directed to undertake the work Commis: Ion ha1 concluded that the

. codify the existing LER reportin8 necessary to develop and demonstrate impact of thb rub will be no greater requirements. See SECY 81-494 for such a cost. effective integrated system than the impa-t of te =xisting LER additional details concerning IOERS. of operational data collection and requirements, emd this rule will not However, the Commission wishes to analyses. place an unaceptable burden on the mike it explicitly clear that it is relaxing If the desi ths reporting requirements with the d'**** '*'"gn of the system affected licenseec, 6

- ch * *Y*
expectation that sulficient utility participation, cooperation, and support easible and cast. effective, development Relotionship Betwsen the LER Rule i

(f 50.73) and the immediate Notification .of the NPRD system will be forthcoming. pkted b Rule (f 50.72) If the NPRD system does not become hould y1 opIrational at a satisfactory levelin a IIL Analysis c: Comments As a parallel activity to the b preparation of I 5053, the Commission ' bo lhhorm of ad ne Commission received forty seven is amending its regulations (l 50.72) issi n on (47) letters commenting on the proposed which require that licensees for nuclear rul$h",*[, bM"the y[ ' rule. Copies of those letters and a power plants notify the NRC Operations published an advanced notice (40 FR detailed analysis of the comments are Center of significant events that occur at 49134) that deferred development of the available for public inspection and their plants. On December 21.1981, the IOER system and sought public copying for a fee at the NRC Public Commission publishedin the Federal comment on the scope and content cf Document Room at 1717 H Street. NW., Register a proposed rule (48 FR 61894) the LER system. Six comment letter'. Washington. D.C. A number of the more that described the planned changes in wers received in response to this substantive issues are discussed below. 5 50J2. ANPRM. All of the comments recched !.!censee Resources De Federal Register notice werereviewed by the staff and were accompanying the proposed LER rule considered in the development of the OQarticular concern to the. changes anticipated to i 50J2 would be (i.e I 5033) stated that additional proposed LER rule. See SECY 82-3 5 for Comrnission was the impact that the additional details, proposed rule would have on the made but they would be "* *

largely l

nls rule identifies the types of resources used by licensees to prepare administrative and the revised i 5052 re:ctor events and problems that are LERs.The Commission's goal was to ' would not be significantly modified nor ( believed to be significant and useful to assure thatlhe scope of the rule would would it be published again for public i the NRC in its effort to identify and not increase the overalllevel of effort comment." Several commenters i resolve threats to public safety. It is above that currently required to comply disagreed with this conclusion. designed to provide the informstion with the existing LER requirements. The commenters did, however, agree necessary for engineering studies of Thirty letters of the 47 received with the Commission's position that l operational anomalies and trends and contained comments on the overall inconsistencies and overlapping patterns analysis of operational acceptability of the proposed rule or requirements between the two rules i occurrences. ne same information can commented directly on the. question of need to be eliminated. aln be used for chr analytic scope and/or resources associated with The Commission has cardully proc idurer that will aid in identifying the proposed rule.De views of the reviewed the proposed requh tments in accident pacursors. commenters can be characterized as the LER and immediate Notification The Commission believes that the follows: rules and has concluded tha'. although NRC should continue to seek an

1. Five commenters felt that the scope changes to both have been made improved operational data system that andlevel of effort would be greatly (largely in response to public comments) willmaximize the value of operational expanded by the proposed rule.

to clarify the intent of the rules, the data.The system should encompass and Estimates included an increase of100 original intent and scope have not been integrate operational data of events and ' man. years for the entire industry, an significantly changed. Therefore, the problem sequences identified in this increase of three times the current effort. Commission has concluded that these rule. NpRDS data. and such other and an increase of $100.000 an'd 2 man-two rule: r.eed act be published again information as is required for a years annually for each plant. for public comment.

33852 Federal Regi:ter / Vol. 48. No.'144 / Tutaday. July 28, 1983 / Rules cnd Regulati:ns Engineering /udgment 'IERs).%ey noted that reports of RPS this rule but did not change the original ~ In the Federal Registar notice that actuadons are already reported to the scope ofintent of the requirements. In NRC in the Monthly Operating Status addition. in order to make the accompanied the proposed rule, the Report as well as telephoned to the requirements in il 50.72 and 50J3 more Commission stated that licensee's NRC Operations Center. compatible, the order (i.e numbering) of engineering judgment may be used to in addition, the Institute of Nuclear the criteria in i 50.73 has been changed. 4 decide if an event is reportable. Several power Operations (INPO) analyzed the ne changes are noted in the discussion commenters expressed the belief that frequency f reactor scrams during a of each paragraph below, some wording should be added to the nemond pedoms analysts Finally, conforming amendments am i rule of reflect that the NRC will also use indicated that an average of 55 reactor being made to various sections ofParts i judgment in enforcement of this trips w uld be reportable each month 20 and 50 in order to reduce the regulation where the licensee is under the proposed rule. INPO equated redundancy in reporting requirements requested to use engineenng dgment this to 600 additionalIIRs per year for ' that apply to operating nuclear power %e Commission believes at the IIR all cunently operating plants, or plants. In general, these amendments rule adequately discusses the need for approdmWy 32 mangean of whquin 6at and application of the concept of - additional effort for all the currently

1. Licensees that have an Emergency

" engineering judgment." The concept perating plants based upon the Notification System (ENS) make the itselfinclude: the recognition of the assumption that each IIR requires 100 reports required by the subject sections existence of a reasonable range of e rt papan and via the ENS. All other licensus will interpretation regardmg this rule, and may n i re and he b owledges' tua inclu reac r

~

the need for flexibility in enforcement freq tl sociated with ss on e eve tha t$dr pt is N sujecta ens be a tied to t o n ety N cument CoMe sufficiently clear and that additional significance. In additio'n. If the ESFs are explicit guidance is not necessary. being challenged during routine g,"*R nalO p Reporting Schedule transients, that fact is of safety'

3. Holders oflicenses to operate a nuclear power plant submit the written in the Federal Register notice that du e

doe not accompanied the proposed rule, the. - agree with the estimate that e'ach LER reports required,by the sub}ect sections in accordance with the procedures Commission stated that it had not yet submitted for a routine reactor trip described in i 50.73(b). decided if the reports should be would require, on the average.100 man. %e crHeria contained in the subket submitted in fifteen days or thirty days hours to prepare and analyze. Licensees sections which define a reportable event following discovery of a reportable are shady required to make internal have not been modified. event. Many commenters stated that the evaluation of and document significant Similar changes are also planned as time fraue for reporting LERs should not. events, including reactor trips. be less th,n thtr+y days after the Therefore, the incrementalimpact of part of curent ectivities to make more discovery of a reportable event. preparing and analyzing the LER should substantive changes to Part 21 One commenter estimated the impact be significantly less than 100-man hours. I 50.55(e), and i 73.71. of a requirement to submit a report in addition, the actualincrease in yonconsefvotive Inte1 dependence sooner than 30 days following discovery burden would be offset by reductions in Several commenters expressed of a reportable event would be an the burden of reporting less significant increase of approximately 40 man years events that would no longer be difficulty in snderstanding the meaning per year for the currently operatin8 reportable. of the phrase "nonconservative plants. In addition the commenter interdependence" as used in the estimated that if a summary report were Coordinarlon With OtherReporting proposed i 50.73(a)(3). The wording of also required the reporting burden Requuwments i 50.73(a)(3)(i 50J3(a)(2)(vii) of this would increase an additional 12 man Several commenters notedthat the final rule) has been changed to eliminate years for the currently operating plants. proposed rule did not appear to be the phrase "non conservative in response to these comments, the coordinated with other existing interdependence" by specifically Commission has decided to require that reporting requirements, and that defining the types of events that should LERs be submitted within 30 days of duplication oflicensee effort might be reported. The revised paragraph does discovery of a reportable event or result. They recommended that LER not, however, change the intent of the situation. reporting be consolidated to eliminate original paragraph. Reporting of Reactar Trips

"f *th" **I*

"8 Sabotage and Thwats of Violence P* P n Section 50.73(a)(1) of the proposed 'ite Commisalon has reviewed Several commenten noted that the rule (i 50.73(a)(2)(iv) of the final rule) s existing NRC reporting requirements. security-related reporting requirements required reporting of any event which (e.g 10 CFR Parts 20 and 21. I 50.55(e), of I 5058(s)(6)(i 50.73(a)(2)(ill) of this results in an unplanned manual or i 50.72. 5 50.73, 1 73.71, and NUREG final rule)) were already contained in automatic actuation of any Engineered 0654) gndhas attempted, to the extent greater detailin 10 CFR 73J1. For Safety Feature (ESF) including the - practicable, to eliminate redundant instance 173J1 requires an act of Reactor Protection System (RPS). Many reporting and to ensure that the various sabotage to be r'eported immediately, commenters agreed that these events reportingreqhlrements are consistent. followed by a written report within is should be trended and analyzed. but hiany of the changes ir. the final LER days.The proposed rule would have disagreed that they deserve to be rule are as a ree-!t of this effort.These required an LER to be filed within 30 singled out as events of special changes resulted in extensive revisions days. Although distribution of reports is sigtificance (i e.. events reportable as in the wording of criter!e centained in - somewhat different, redundant reporting

s Fedred Registre / Vol. 48. No.144 / Tuesday. July 28, 1983 / Rules and Regulations 33853 would have occurred.%: commenters in NPRDS as an alternative. it is'our Severalcpmmenters argued that the recornmended that the Cc mmission understanding, however, the NPRDS will inclusion of the requiremsnt that the ensure consistency boxeen il 50.73 soon adopt the EHS system titles, so a licensee perform an engineering and 73.71. distinction should no longer exist. In. evaluation of certair. events at the staffs In response to these comments the addition. IIRs frequently include request appeared un,ustified and would Commission has deleted the reporting of systems that are not included in the add substantially te 'he burden of szbotage and threats of violence from scope of NpRDS (i.e an NpRDS systIm 3 porting.They arged that the licensee 150.73 because these situations are identification does not exist) while EHS. s'nould be required (o submit only the adequately covered by the reporting on the other hand, includes all of the snecific additiona: information required requirements contained in 173.71. systems commonly found in commercial io-the necessary engineering evaluation Evocuation ofRooms orb'uildings udeat pown plants. Furan. NpRDS rahr than to pedmm We esakauan includes only 39 componer t identifiers ne rule has been modified to require Many commenters stated that the (e.g valve, pump).ne Coran6sion only the submittal of any necessary reporting ofin. plant releases of, believes that this limited number does additionalinformation requested by the radioactivity that require evacuation of not provide a sufficiently details d Commission in writing / individual rooms (l 50.73(a)(7) in the description of the component fusction proposed rule or (l 50.73(s)(2)(x) of this invelved. IV. Specific Findings final rule) was inconsistent wuh the ge'neral thrust of the rule to require Function ofFoiled Components and Overview of the LER System reporting of significant events. ney Stotus ofRedundant Compor ents When this finalIIR rule becomes n:ted that minor spills, small gaseous Mary commenters said that effective. the IIR will be a detailed waste releases, or the disturbance of inforn.ation rertuired in (150.73(b)(2) (vi) narrative description of potentially contaminated particulate matter (e.g., an'd (vil) of the proposed rule should not significant safety events. By describing du..) may all require the temporary be a requirement in the IIR.%ey in detail the event and the planned evruation of Individual rooms until the argued that this information is readily corrective action,it will provide the t.ltbme concentrations decrease or available in documents previously basis for the careful study of events or until respiratory protection devices are submitted to the NRC by licensees and conditions tha't might lead to serious utilized, ney noted that these events are available for reference. accidents. lf the NRC staff decides that cre fairly common and shouM not be ne final rule (i 50.73(b)(2)(1)(G)) has the event was especially significant reportable unless the required been modified to narrow the scope of from the standpoint of safety, the staff ev:cuation aff ts the entire facility or a the information requested by the may request that the licensee provide use to5: comments the additionalinformation and data Commission. wording of this criterion (i 50.73(a)(2)(x) While this generalinfonnation may be associated W 6e mat. In the Onal rule) has been changed to available in licensee documents ne licensee will prepare an f.rR for signitcantly narrow the scope of the previously submitted to the NRC. the those events or conditions that meet one etherion to include only those events Commission believes that a general or more of the criteria contained in which sI nificantly hamper the ability of understanding of the event andits I 50.73(a).ne criteria are based 8 sita personnel to perform safety-related significance should be pos:1ble without primarily on the nature, course, and refe' ence to additional. documentation consequences of fhe event,%erefore. r activities (e.g. evacuation of the main control room). which may not be readilv or widely the finalIIR rule requires that events available.particularly te the public. which meet tha criteria are to be Energy Industry Identification System he Commission contkues to believe reported regardless of the plant Many commenters noted that the that the licensee should prepare an IIR operating mode or power level,and requirement to report the Energy in sufficientdepth so that regardless of the safety =Ignmeince of Industry identification System (EHS) knowledgeable readers who are - the components, systems, or structures involved. In trying to develop criteria for componentfunctionidentiderand 1 conversant with the design of 'ts, butthe identification of events reportable as system name of each component or commercial nuclear power plan system referred to in the LER -q are not familiar with the details of a. LERs. the Commission has concentrated description wouldbe a alonmrint particularplant, can understand the on the potentialconsequences of the burden on thelicensee. ~ general character'stics of the event (e.g., eventas the measure of significance. Hey suggested instead that the the cause,the synmeine. the. %erefore, the reporting criteria. in NpRDS component identifiers be used in corrective action). As. suggested bythe general, do not specifically address place of the EIIS component identifiers commenters, more detailed information classes of initiating events or causes of which are not yet widely used by the to support engineering evaluations and the event. For example there is Ri ' industry. case studies willbe obtained.as requirement that all personnel errors be na Commission continues to believe needed, directly from the previously reported. However. many reportable that EDS system names and component submitted licenses documents. events willinvolve or have been function identifiers are needed in order Initiated by personnel arrors. Engineerirg Bruleat/arrs-Finally.it should be noted that that IIRs from different plants can be compared. We do not, however. suggest %e overview discussion of the licensees are permitted and encouraged thaJ the EDS identifiers be used proposed rule contains the following to report'any event that does not meet throughout the plant but only that they statement:"If the NRC staff decides that the criteria containedin i 50.73(a),if the be added to the LER as itis written. A the event was especially significant licensee believes that the event might be simple, inexpensive table cottld be used-from the standpoint of safety the staff of safety significance, or of generic to translate plant identifiers into may request that the licensee perform interest or concern. Reporting equivalent EUS identifiers. an engineering evaluation of the event requirements aside assurance of safe The Commission considered the and describe the results of that operation of all plants depends on system and component identihrs used evaluation." accurate and complete reporting by each

1 33854 Feder:1 Regist:r / Vol. 48. No.144 / Tuisd:y. July 28. 1983 / Ruiss and Regulati:ns licensee of all events having potential consequ'ences of an event (e.g at the accident (e.g., containment isolation. safety significance, discretion of thelicensee as part of a emergency filtration). Hence. minor Parugmph.by.Pomsmph Explanation of planned procedure or evolution). operational events involving a specift, the LER Rule Sections 50.73(a (2)(v) and (vi) component such as valve packing teaks. (proposed l 50.73ga)(2)) require reporting wnica could te considerd a lack af De significant provisions of the final of: control of radioe:tive material, should ' !?A rule are explained below.The not be reported uader this parapaph. i explanation follows the order in the (v) Any event or condition that alone could System leaks or ot'ter similar eve 'its y pjposed rule. have prevented the fulfillment of the safety may, however, be reportable under other aragraph 50.73(a)(2)(lv) (proposed function of stractures or systema that are i pargraph 50.73(a)(1)) requires reporting needed to:' paragraphs, It should be noted that there afe a. of:,,Any event or condition that resulted (A) Shut down b reactor and maintain it limited number of single. train systems in manual or automatic actuation of any la a safe shutdown condition: that perform safety functions (e.g4 the. Engineered Safety Feature (ESF). (B) Remove residual host: Including the Reactor Protection System (c) control the release of radioactive High Pressure Coolant lajection System ra systems.bse oW matedah or. si gle train would prevent the .(RPS). However, actuation of an ESF sete 6e e neequences of an including the RPS. that resulted from accident. fulMment of the safety function'of that and W88 part of the preplanned,. (vi) Events covered in parasreph (a)(2)(v) systen and therefore, must be reported sequence during testing or reactor. of this section may include one or more even though the plant Technical operation need not be reported. personnel errore, equipment failures. and/or Speclicatons may allow such a This paragraph requires events to be. discovery of deelsn. analysis fabrication. ' con /ltion to exist for a ap*~CiBed lim.ited reported whenever an ESF actuates construction, and/or procedural le@ he. n eithermanuall orautomatically, inadequacles.However. Individual It should also be noted that,if a S regardless of pfant status. It is based on.. " failures need not be reported. l the premise that the ESFs are provided pursuant to this paragraph if redundant potentially serious human error is made to mitigate the consequences of a equipment in se same system was operable, that could have prevented fulMment of and available to perform the required safety a safety function, but recovery factors. significant event and. therefore: (1) %ey should work propetly when called upon. resulted in the error being corrected. the' and (2) they snould not be challenged .been changed from the proposed rule to _ error is still report %e wording of this' paragraph has %e Commission recognizes that the frequently or unnecessarily.The Commission is interested both in events make it easier to read.%e intent and application of this and other paragraphs where an ESF was needed to mitigste scope of the paragraph have not been of this section involves the use of the consequences (whether or not the changed. engineering judgment on the part of equipment performed properly) sad De intent of this paragraph is to licensees.In this case, a technical events where an ESF operated capture those events where there would judgment must be made whether a unnecessardy, have been a failure of a safety system to failure or operator action that did " Actuation" of multichanne? ESF properly complete a safety function, cetually disable one train of a safety Actuation Systems is defined as regardless of when the failures were sptem. could have, but did not, affect a actuation of enough channels to discovered or whether the system was reemdant train within the ESF system. complete the minimum actuation logic needed at the time. If so this would constitute an event that (i.e., activation of sufficient channels to his paragraph is also based on the "could have prevented" the fulfillment cause activation of the ESF Actuation assumption that safety.related systems of a safety function, and. accordingly. System).Therefore. single channel and. structures are intended to mitigate must be reported. actuations, whether caused by failures the consequences of an accident. While if a component falls by an apparently or otherwise, are not reportable if they i 50J3(a)(2)(lv) of this final rule applies random mechanism it may or may not do not complete the minimum actuation to actual actuations of an ESF. be reportable if the functionally logic. 150J3(a)(2)(v) of this final rule covers redundant component could fall by the Operation of an ESF as part of a an event or condition where redundant same mechanism. Reporting is required planned operational procedure or test structures. components, or trains of a if the failure constitutes a condition (e.g., startup testing) need not be safety system could have failed to whern there is reasonable doubt that the reported,liowever,if during the planned perform their intended function because functionally redundant train or channel operating procedure or test, the ESP of: one or more personnel errors, would nmain operational untilit actuates in a way that is not part of the including procedure violations: completed its safety function or is planned procedure, that actuation must ^ equipment failures or design, analysis, repaired. For example,if a pump in one be reported. For example,if the normal fabrication, construction, or procedural train of an ESF system fails because of reactor shutdown procedure requires deficiencies. %e event must be reported improper lubrication, and engineering that the control rode be inserted by a regardless of the situation or condition judgment indicates that there is a ' manual reactor trip, the reactor trip need that caused the. structure or systems to reasonable expectation that the not be reported.14 wever, if conditions be unavailable, and regardless of functionally redundant pump in the develop during the ibutdown that whether or not an alternate safety other train, which was also improperly require an automatic scactor trip, such a. system could have been used to perform lubricated, would have also failed reador trip must be repet, the safety function (e.g High Pressure before it completed its safety function. The fact that the safety analysis Core Cooling failed but feed.and-bleed then the actual failure is reportable,and assumes that an ESF will actuate orlow Pressure Core Cooling were the potential failure of the functionally automatically during certain plant available't6 pbvide the safety function redundant pump must be discussed in conditions does not eliminate the need obcore cooling). the I.ER. to report that actuation. Actuations that The applicability of this paragraph For safety systems that include three need not be reported are those initiated includes those safety systems designed or more trains. the failure of two or more for reasons other than to mitigste the to mitigate the consequences of an ' trains should be reported if. In the

Fcdtral Register / Vol. 48. No.144 / Tuted:y, July 28. 1983 / Ru.es and Regulations 33855 ~ judg: ment of the licensee, the hmctional (D) Mitigate the consequences of an within the time limit specified in the captbility of the overall system was accident." Technical Specifications, the action leopardized. This paragraph has been changed t5 ceed not be reported under this Interaction between systems; clarify the intent of tne phrase, paragraph. However. if. while the train p:rticularly a safety system and a non. "nonconservative interdepen lence/f or component is out of service the s:faty system. is also included in this Numerous comment letters expressed licensee identifies a condition that could critzrfon. For example, the Commission difficulty in understanding what this have prevented the whole system from is increasingly concerned about the phrase meant:so the paragraph has performing its intended function (e.g, effect of a loss or degradation of what been changed to be more specific.The the licensee finds a set of relays that is had been assumed to be non-essential new paragraph is narrower in scope wired incorrectly), that condition must inputs to safety systems.Therefore, this than the original paragraph because the be reported. par: graph also includes those cases term is specifically defined, but the Section 50.73(a)(2)(1)(proposed whIre a service (e.g., beating. bade intent is the same. 150.73(a)(4)) requires reporting of: } ventilation, and cooling) or input (e.g Es paragraph requins those events "(A)The completionof anynuclear compressed air) which is necessary for - to be. sported where a single cause plant shutdown required by the plant's reliable orlong-term operation of a produced a component or group of Techical Sp&h w ' safsty system is lost or degraded. Such components to become inoperable in , loss er degradation is repo. table if the redundant or independent portions (i.e "(B) Any operation prohibited by the WedMmw P propIr fulfiument of the safoty function trains or channels) of one or more p is not cannot be assured. Failures that systems having a safety function.%ese "(C) Any deviation from the plant,s affect inputs or services to systems that events can identify previously Technical Specifications authorized h:ve no safety function need not be unrecognized common cause failures pursuant to i 50.54(x) of this part., nported. - and systems lateractions. Such failures This paragraph has been reworded to can be simultaneous failures which more clearly define the tvents that must Finally the Commission recognizes that the licensee may also use ccur because of a single initiating be reported. In addition, the scope has engmeering judgment to decide when cause (Le., the single cause or been changed to requ!re the reporting of personnel actions could have prevented mechanism serves as a common input to events or conditione " prohibited by the fulfillment of a safety function.For the fauunsh w the faums can be plant's Technical Specifica, tons" ra ther exanple, when an individual improperly sequential (i.e., cascade failures), such than events where "a plant Technical oper.-tes or maintains a component, he as the case where a single component Specification Action Statement is not failure results in the failure of one o" met."nis change accommodates plants might conceivably have made the same errorforallof thefunctionally more additionalcomponents. that do not have requirements that are redundant components (e.g,if he To be repoytable, however, the event specificallydefined as Action r failure most result in or involve the Statements. incorrectly calibrates one bistable emplifier in the Reactor Protection failure findependent portions of more his paragraph now requires events to than one train a channelin the same " System.he could conceivably be reported where the licensee is inco'rrectly calibrate all bistable WHennt sistems. Fw example. U a required to shut down the plant because amplifiers). However. for an event to be

- "'"".mnedon caused components the requirements of the-Technical reportable it is necessary that the in Train A and 'B" of a single system Specifications were not met.For the to become inoperable, even if additional purpose of this paragraph.." shutdown" actions actunuy affect or involve trains (e.g, Train. C") were still is defined as the point in time where the components inmore than one train or chamel of a safety system, and the available, the event must.be reported. In TechnicalSpecifications require that the addition. If the cause or conditio result of the actions must be undesirable camd umponents in Train,A,,n plant be in the first shutdown condition ofone required by a Ilmiting Condition for from the perspective of protecting the -

smem andin nain Wander h5alth and safety of the public. no Operation (e.g hot standby (Mode 3) for tr components can be functionally '[* 1",gy with the StandarMechnM t indep den ) to Speedications). E the condition is redundant (e g., two pumps in different b trains) or not functionally redundant corrected before the time limit for being rep n a ca shut down (Le before completion of the (e.g, the operator correctly stops a pump in Train "A" and, instead of shutting the condition caused com[ Train "A"of shutdown), the event need not be nents in Train. "A" of one systern an pump discharge valve in Train "A." he another system (Le., trains that are not nponed. miatakenly shuts the pump dischar8" assumed in the safety analysis to be In addition. if a condition that war valve in Train "B"). independent), the event need not be prohibited by the Technical Section 50.73(a)(2)(vil)(proposed reported unless it meets one ormore of . Specifications existed for a peded of I 50.73(a)(3)) requires the reporting of: the other criteria in this section. time longer than that permitted by the "Any event where a single cause or in addition, this paragraph does not Technical. Specifications. it must be condition caused at least one include those cases where one train of a reported even if the condition was not independent tr:!a of channel to become system or a component was removed discovered until after the allowable time inoperable in multiple systems or two from service as part of a planned had elapsed and the condition was indipendent trUns channels or to evolution,in accordance with an rectified immediately after discovery. become inoperable in a systern designed approved procedure, and in accordance Section 50.73(a)(2)(II)(proposed to: with the plant's Technical - 150.73(a)(5)) requires reporting of:"Any (A) Shut down the reactor and Specifications. For example,if the - event or condition that resulted in the m:intain it in a safe shutdown licensee removes part of a system from condition of the nuclearpowerplant, condition. service to perform maintenance, and the inclu' ding its principal safetybarriers. (B) Remove residual heet. Technical Specifications permit the being seriously degraded, or that j (C) Control the release of radioactive resulting configuration, and the system resulted in the nuclestpower plant material: or or component is returned to service being-

33856 Ndsral Registre / Vol. 48. No.144 / Tuesday. July 28, 1983 / Ruls: and'Regulatitns ~ "(A) In an unanalyzed condition that radioactivity levels at a BWR air ejector safety of the nuclear power plant or sign:ficantly compromised plant safety; monitor that exceeded the Technical significandy hampered site personnella "(b)in a condit'on that was outside Specification limita. the performance of duties necessary for the design basis of the plant:or (c) Cracks and breaks in piping, the the safe operation of the nuclear power i I "(C) In a co.dition not covered by the reactor vessel, or major components in plant including fires, toxic gas releases. plant's operating and emergency the primary coolant circuit that have or radioactive releases." procedures." safety relevance (steam generators. This paragraph has been reworded to This paragraph requires events to be reactor coolant pumps, valves, etc.) include physical hazards (Internal to the reported where the plant, including its (d) SignTicant welding or material plant) to personnel (e.g., electrical Ares). principal safety barriers, was seriously defects in tre primary coolant system. In addition. In response to numerous degraded or in an unanalyzed condition. (e) Scrio.is temperature or pressure comments, the scope has been narrowed For example, small voids in systems - transients (e.g transients that violate so that the hazard must hamper the designed to remove heat from the the plant's Technical Specifications). ability of site penonnel to perfonn - reactor core which have been previously (f) Loss of relief and/or safety valve safety-related activities affecting plant shen through analysis not to be safety operability during test or operation-safety' signifient need not be reported. (such that the number of operable In-plant releases must be reported if However, the accumulation of voids that valves or man way closures is less than they require nacuadon of rooms or could inhibit the ability to adequately required by t, a Technical buildings containing systems important remove heat from the reactor core. Specifications).. to safety and, as a result, the ability of, particularly under natural cirealation . (g) Loss of containment function or coriditiona, would constitute an integrity (spcontainment leakage rates the operators to perform necessary,,, safety functions is signincandy h unanalyzed condition and must be exceeding tne authorized'11mits). hampered. Precautionary evacusticesj reported. In addition, voiding in Section 50.73(a)(2)(lii) (proposed instrument lines that results in an i 50.73(a)(0)) requires reporting of: "Any roorns and buildings that subsequent. - erroneous ind! code causing the natural phenomenon or other extemal evaluation determines were not required operator to significan'ly misunderstand condition that posed an actual threat to need not be reported. i the true condition of Q,e plant is alsotan the safety of the nuclear power plant or' Proposed i 50.73(a)(8) was latended to unanalyzed conditiw and must be significantly hampered site personnelin capture an event that involved a reported. the performance of duties necessary for controlled release of a signi8 cant The Commission recognizes that the the safe operatiori of the nuclear power am6unt of radioactive material to offsite i licensee may use engineering judgment plant." areas. In addition. "signi8 cant" was and experience to determine whether an his paragraph has been reworded to based on the plant's Technical unanalyzed condition existed. It is not make it clear that it applies only to acts SpeciScation ilmits for the release of intended that this paragraph apply to of nature (e.g, tornadoes) and external radioactive material. However, this l minor variations in individual haznrds (e.g, railroad tank car section has been deleted because the parameters, or to problems conceming explosion). References to acts of reporting of these events is already single pieces of equipment. For example, sabatsge have been removed because required by 5 50.73(a)(2)(1) arid i 20.405. at any time, one,or more safety related they :re covered by $ 73.7L In addition. Section 50.73(a)(2)(viii) and (ix) components may be out of service due threats to personnel from internal (proposed i 50.73(a)(9)) require reporting to testing, maintenance, or a fault that hazards te.g radioactivity releases) are of: has not yet been repaired. Any trivial now covered by a separate paragraph single failure or minor error in (i 50.73(2)(2)(x)). (vill)(A) Any airbome radioactivity release performing surveillance tests could This paragraph requires those events that exceeded 2 times the applicable produce a situation in which tw, or to be reported where there is an actual concentrations of the limits specirled in Table more often unrelated, safety related threat to the plant from an external II of Appendix B to part 20 of this chapter la components are out-et.se;vice. condition or natural phenomenon. and unrestricted arus, when averaged over a Technically, this is an unanalyzed where the threat or damage challenges time period of one hour, condition. However, these events should the ability of the plant to continue to (B) Any liquid effluent release that be reported only if they involve operate in a safe manner (including the exceeded 2 timu the limiting combined Maximum Penniselble Concentration (MPC) functionally related components or if orderly shutdown and maintenance of $',',(*j', $ ^ y h,B kohhis they significantiv compromise plant shutdown conditions. safety. Thelicenseeis to ecideif a receiving water (i.e unrestricted arealfor at , Finally, this paragraph also includes phenomenon or condition actually radionucildes except tritium and dissolved material (e.g., metallurgical chemical) threatened the plant. For example, a noble sun, when averaged over e time problems that cause abnormal minorbrush fire in a remote area of the period of one hour. degradation of the principal safety site that was quickly controlled by fire (lx) Reports submitted to the Co==t a barriers (i.e., the fuel cladding, reactor fighting personnel and, as a result, did in accordance with peregre (aX3Xvill]'ef thle petion also mut the e uent release coolant system pressure boundary, or not present a threat to the plant need reporting requirements of paragraph the containment). not be reported. However, a major forest to M eg ono 20 of this chapter, Additional examples of situations fire.large scale flood, or major included in this paragraph are: earthquake that presents a clear threat (a) Fuel cladding failures in the to the plant must be reported. Industrial Paragraph (vill)has been changed to reactor or in the ' storage poo'. that or transportation accidents that clarify thelequirements to report exceed expected values, ti:st are unique oca:urred near the site and created a releases of radioactive matedal.'line or widespread, or that resulted from plant saf6ty dbncem must also be paragraph is similar to i 20.408 but. unexpected factors. rported. places a lower threshold for reporting. (b) Reactor coolant radioactMty Section 50.73(a)(2)(x) (proposed events at commercial power reactors, levels that exceeded Technical I 50.73(a)(7)) requires reporting of:"Any ne lower thresholdis based on the Specification limits for lodine spikes or, event that posed an actual threat to the significance of the breakdown of tle ,3

Fed rd Re'ist:r / Vol. 48. N:.144 / Tu:sdty July 26. 1983 / RuW and Regulstions 33857 g lic:nsee's program necessary to have a in a condition not analyzed in the Safety "Special Reports" of the Technical

release of this site, rather than on the Analysis Report) under reasonable a nd Specifications are still required.

significance of the impact of the actual credible alternative conditions such as y,g,,yj,,,,y go,;y,;, release. power level or operaung mode. For Reports of events covered by exaraple.'if an event occurr'edMhile t}p The Commission has prepared a i 50.73(a)(2)(viii) are to be made in lieu plant was at15% power and the same regulatory analysis for t. 's final rule. of reporting noble gas releases that event could have occurred while the The analysis examines one costs and exceed 10 times the instantaneous plant was at 100% power and. as a benefits of the alternatives considered i release rate, without aversging over a result, the consequences would have by the Commission. A copy of the time period, as implied by the been considerably more serious, the regulatory analysis is andable for r:quirement of I 20.405(e)(5). IIcensee must assess and report those inspection and copying for a fee at the Paragraph 50.73(b) describes the ' consequences. NRC Public Document Room.1717 H l format and content of the IIR. It Paragraph 50.73(b)(4) requires that the Street. N.W Washington, D.C. Single requires that the licensee prepare the licensee describe in the LER any copies of the analysis may be obtained LER in sufficient depth so that corrective actions planned as a result of from Frededek J. Hebon. Chief. Program kn:wledgeable readers conversant with the event that are known at the time the Technology Branch. Office for Analysis I the design of commercial nuclear power LER is submitted. including actions to and Evaluation of Operational Data, plants, but not familiar with the details reduce the probability of similar events U.S. Nuclear Regulatory Commission, of a particular plant, can understand the occuidng in the future. After the initial Washington. D.C. 20555: Telephone (301) complete event (i.e., the cause of the LER la sut,mitted only substantial 492-4480. event, the plant status before the event, changes in the corrective action need be VL Paperwork Reduction Act Statement and the sequence of occurrences during reported as a supplemental LER. O' *?'"II' h 5073(b)(1) requires that the Paragraph 5053(c) authorizes the NRC ne Nuclear Regulatory Commission Paragrap staff to require the licensee to submit has submitted this rule to the Office of lic nsee provide a brief abstract specific supplementalinformation Management and Budget for such descdbing the major occurrences dudng beyond that required by I 50J3(b). Such review as may be appropriata under the thievent. including all actual Paperwork Reduction Act. Pub.l.06-component or system failures that information may be relutred if the staff511.The date on which the reporting contributed to the event, all relevant finds that supplementa materialis necessary for complete understanding of requirements of this rule become - .cp;rator errors or violations of an unusually complex or significant effective reflects inclusion of the 60-day proc dures,and an significant corrective action en or planned as a event. Such requests for supplemental period which the Act allows for such result of the event.This paragraph is infadon mut h mdhd% review. ' needid to give LER data base users a and the licensee must submit the bl. Regulatory Flaxibility Castification brief description of the event in order to requested information as a supplement to the initial LER within a time period in accordance with the Regulato mu y agreed upon by the NRC staff F exi i t of 1 5 C. ( ), Parag ph 3 2 uires that the n rbe act on u tia h4 Ip ne e ti vent' Exe ti Dire or e ati a the so that mders not familiar with the authority to grant case-by-case number of small entities. This final rule details oh 3 articular plant can exemptions to the reporting affets electric dtigs that are undi rstand 'he event.The licensee requirements contained in the LER dominant in their respective service sh:u d imphi size how systems, system.This exemption could be used to areas and that own and operate nuclear utilization facilities licensed under ~comoor ents, ud operating personnel limit the collection of certain data in pedort ted. Spsofic hardware problems those cases where full participation. sections 103 and 104b of the Atomic rhuld nct be covered in excessive would be unduly difficult because of a Energy Act of 1954, as amended.De det:ll. Characteristics of a plant that are plant's unique design or circumstances, amendments clarify and modify . unique and that influenced the event Paragraph 50J3(g) states that the presently. existing notification (favorably or unfavorably) mus~ be reporting requirements contained in requiremets. t d: scribed.The narrative must also i 50J3 replace the reporting Accordingly, there la no new,, describe the event from the perspective requirements in all nuclear power plant significant economic impact on these Technical Specifications that are licensees. nor do these licensees fall ef tha operator (e.g., what the o[erstor saw, did. perceived, understoo or typically associated with Reportable within the scope of the definition of misunderstood). Occurrences. "small entitles" set forth in the Paragraph 50J3(b)(3) requires that the ne reporting' requirements Regulatory Flexibility Act or the Small ^ LER include a summary assessment of superseded by I 50J3 are those Business Size Standards set out in l the actual and potential safety contained in the Technical Specification regulations lasued by the Small Business consequences andimplications of the sections that are usually titled " Prompt Administration at 13 CFR Part 121. event.nts assessment may be based on Notification with Written Followup" List of Subjects the conditions existing at the time of the (Section 6.9.1.8) and

Thirty Day Written-m CM PorUo 7

event.The evaltiation must be carried Reports" (Section 0.9.1.9). %e reporting cut to the extent necessary to fully requirements that have been superseded Licensed material. Nuclear power assess the safety consequences and are also described in Regulatory Guide plants and reactors. Penalty, Reporttag safety margins associated with the 1.16, Revision 4. " Reporting of Operating and recordkeepin8 requirements. ev:nt. An assessment of the event under Information-Appendix ATechnical alternative conditions must be included Specification." Paragraph 2. " Reportable. M CFR PARTS $0 if the incident would have been more Occurrences." The special report incorporation by reference, Antitrust. severe (e.g4 the plant would have been typically described in Secticn 6.94 Classified information. Fire protwtion.

33858 Fed:ral Rogist:r'/ Vol. 48, No.144 / Tuesday, July 28, 1983 f Rul:s cud R2gulati:ns Intergovernmental relations Nuclear (C) Any deviation from the plant's (D) Mitigste the consequences of an power plants and reactors. Penalty. Technical Specifications authorized accident. Radiation protection. Reporting and pursuant to i 50.54(x) of this part. (viii)(A) Any airborne radioactivity recordkeeping requirements." (ii) Any event or condition that release that exceeded 2 times the resulted in the condition of the nuclear applicable concentrations of the limits Under the authority of the Atomic Energy Act of 1954, as amended, the power plant. Including its principal specified in Appendix B, Table 11 of Part Energy Reorgamzation Act of 1974, as safety barrien, bemg seriously 20 of this chapter in unrestricted areas, amended, and 5 U.S C. 552 and 553, the degraded, or that resulted in the nuclear when averaged over a time period of f6110 wing amendments to 10 CFR Parts p wer plant being: one hour. 20 and 50 are published as a document (A)In an unanalyad condition that (B) Any liquid effluent release that subject to codification. significantly compromteed phnt safety; exceeded 2 times the limiting combined l (B)In a condition that s as outside the Maximum permissible Concentration. (MPC)(see Note 1 of Appendix B to Part PART 50-DOMESTIC UCENSING OF design basis of tha plant or 20 of this chapter) at the point of entry; PRODUCTION AND UT!UZATION (C)In a condition not covered by the FACli3 TIES plant's operating a.1d emergency into the receiving water (Le procedures. unrestricted area) for all radionuclides 1

1. We authority citation fo,r Part 50 (lii) Any natural pheamenon or other except tritlum and dinohed noble continues to read as follows:

^ external condition that peed an actual gases, when averaged over a time period Authority: Sees.103,104.181.182.183,1os. threat to the safety of the nuc: ear power of one hour. 189, sa Stat. 938,937,948,953,954,955, s54, aa plant or aIgnificanJy tiamperel site (1x) Reports submitted to the ur-amended. sec. 234. 83 Stat.1244, as amended personnel in the performance of duties Comminalon in accordance with, i t.b. 1 (42 USC 2133, 2134, 2201. 2232, 2233. 223e; necenery for the safe operation of the paragraph (a)(2)(viillof this section also. 2239. 22s2): seca. 20t. 202. 20s, as Stat.1242. nuclear power plant. meet the effluent release reporting 1244.1246, as amended (42 UAC 5841. 5842. (iv) Any event or condition that requirements of paragraph 20.406(a)(5). 5848), unless otherwise noted. resulted in manual or automatic of Part 20 of this chapter. Section 50J also issued under Pub. L 95-actuation of any Enginee' red Safety (x) Any event that posed an actual + e01. sec.10,92 Stat. 2951 (42 UAC 5851). Feature (ESF), including the Reactor. threat to the safety of the nuclear power Protection System (RPS). However, plant or significantly hampered site un r Pub. 97 5,96 Sta 2 3( 2 .C actuation of an ESF, including the RPS, personnelin the performance of duties 2239). Section 50Js also tasued under sec. 122. ea Stat. 939 (42 UAC 2152). Sections 11 at resulted fmm and was part of the necessary for the safe o tion of the 50.804ost also issued under sec.164,68 Stat. preplanned sequence during testing or nuclear power plant ding fires, 954. as amended (42 USC 2234). Sections nactor operstion need not be reported. toxic gas releases, or radioactive 50.100-50-102 also issued under sec.180, eo (v) ev.y event or condition that alone releases. Stat. 955 (42 UAC 223c). could have prevented the fulfillment of (b) Contents. %e Licensee Event For the purposes of sec.123. 68 Stat. 958, as the safety function of structures or Report shall contain: amended (42 UAC 2273), il 50.10 (a), (b), systems that are needed to: (1) A brief abstract describing the and (c). 50.44. 50.48, 50.48, Sa54, and 50.80(a) (A) Shut down the reactor and major occurrences during the event, are issued under sec.161b. 68 Stat. 946, as maintain it in a :afe shutdown including all component or system d condition: failures that contributed to the event (c and a ic un er sec. tot Stat. 949, as amen <ied (42 UAC 2201(1)); and (B) Remove residual heat: and significant corrective action taken il 50.55(e),50.59(b),50.70,5071. 50.72. and (C) Control the release of radioactive or planned to prevent recurrence. 5038 are issued under sec. telo,68 Stat. 950, material: or (2)(1) A clear, specific. narrative as amended (42 UAC 22a1(o)). (D) Mitigate the consequences of an description of what occurred so that accident. knowledgeable readers conversant with

2. A new I 50.73 is added to read as (vi) Events covered in paragraph the design of commercial nuclear power I UO*8' (a)(2)(v) of this section may include one plants, but not familiar with the details i 50.73 Ucensee event report system.

or more procedural errors, equipment of a particular plant, can understand the failures, and/or discovery of design, complete event. (a) RePonoble events. (t) ne holder analysis, fabrication, construction, and/ (ii) The narrative description must of an operating license for a nuclear or procedural inadequacies. However. Include the following specific power plant (licensee) shall submit a individual component failures need not information as appropriate for the Licensee Event Report (LER) for any be reported pursuant to this paragraph if particular event: event of the type described in this redundant equipment in the same (A) Plant operating conditions before paragraph within 30 days after the system was operable and available to" the event. discovery of the event. Unless otherwise perform the required safety function. (B) Status of structures,com onents, specified in this section, the licensee (vil) Any event where a single cause' ' or systems that were inoperahfe at tho' shall report an event regardless of the or condition caused at least one start of the event and that contributed to plant mode or power level, and independent train or channel to become the event. regardless of the significance of the inoperable in multiple systems or two (C) Dates and approximate times of. structure, system, or component that independent trains oc channels to occurrences. initiated the event. become inoperable in a single system (D) De cause of each component or. (2) The licensee shall report: designed'to: system failure or personnel error. if (I)(A) The completion of any nuclear (A) Shut dowin the reactor and known. plant shutdown required by the plant's maidtain it in a safe shutdown (E)The failure mode, mechanism, and Technical Specifications; or condition: effect of each failed component. if (B) Any operation or condition (D) Remove residual heat: kncwn. prohibited by the plant's Technica.1 (C) Control the release of radioactive (F) The t nirgy ladustry identdication Specifications: or. meterf ah or System compone.t functio".a identifier

Federal Resistsr / Vol. 48. No.144 / Tuesday, July 26.1983 / Rult, and Regulations 33859 and system name of each component or components that could have performed PART 20-STANDARDS FOR system referred to in the LER. the same function as the components PROTECTION AGAINST RADIATION (f)The Energy Industry Identification and systems that failed during the event.

3. In $ 20.402, paragraph (s) is revised:

System is defined in: IEEE Std 803.-1983 (4) A description of any rrective the introductory text of paragraph (b)is (May 18.1983) Recommended Practices actions planned as a result Ith'e evenf. revised: and a new persgraph (e)is for Unique identification Plants and tl the a d Facilitie+-Principles and fg* tc ng in the (2)lEEE iltd 803-1933 has been (5) Reference to any previous similar 12M2 - Reports of men or lose of approved for incorporation by reference. events at the same plant that are known fcenew material to the licensee. (a)(1) Each licensee shall report to the by the Director of the Federal Register, A n tice of any changes made to the (8) %e name and telephone number of Commission by telephone.imtnedlately after it determines that a loss or thef t of materialincorporated by reference will a person within the Ucensee's licensed material has occurred in such be published in the Federal Register. organization who is knowledgeable {uantitles and under such circumstances Copies may be obtained from the about the event and can provide at it appears to the !!censee that a Institute of Electrical and Electronics additional lnformation concerning the substantial hszard may result to persons Enginsers. 345 East 47th Street. New event and the plant's characteristics. Ycrk. NY 10017. A copy is avallable for (c) Supplementallnformation. The (2) Repom mst be mah a hwo inspection and copying for a fee at the Commission may require the licensee to (1) Licensees having an installed ( Commission's Public Document Room. adbmit specific additionalinformation Emergency Notification System shall i 1717 H Street. NW., Washington. D.C. beyond that nquired by paragraph (b) make the reports to the NRC Operations and at the Office of the Federal Register. of this section if the Camission finds Centerin accordance with 150.72 of this 1100 L St. NW, Washington. D.C. that supplemental materialis necessary chapter. (C) For failures of components with multiple functions, include a list of for complete understanding of an (II) All other Ucensees shallmake syst:ms or secondary functions that unusually complex or significant event, reports to the Administrator of the were also affected. Dese requests for supplemental appropriate NRC Regional Office IIst.ed (H) For failure that rendered a train of information will be made in writing and in Appendix D of this part. a safety system inoperable, an estimate the licenses shall submit the requested (b) Each licensee who makes a report of ths elapsed time from the discovery infonnation as a supplement to the under paragraph (a) of this section shall, cf the failure until the train was returned initial LER. withing 30 days after learning of the loss 13s:rvice. (d)Subniss/on ofreports. Licensee or theft, make a report in writing to the (!)he method of discovery of each Event Reports must be prepared on U.S. Nuclear Regulatory Commission, component or system failure or Form NRC306 and submitted within 30 Document Control Desk. Washington, proedural error, days of discovery of a reportable event D.C. 20555, with a copy to the (J)(1) Operator actions that affected or situation to the U.S. Nuclear appropriate NRC Regioned Office listed the course of the event. Including Regulatory Commisalon. Document la Appendix D of this part.The report Control Pesk. Washtngton. D.C. 20555. shall include the following information: erbot t contri ut n The licensee shall alsmbmit an (2) For each personnel error; the IJc:nsee shall discuss: additional copy to the appropriate NRC (e) For holders of an operating license (i) Whether the error was a cognitive RegionalOfficelistedin Appendix A to for a nuclear power plant the events error (e.g, failure to recognize the actual part 73 of this chapter, included in paragraph (b) of this section plant condition. failure to realize which (e) Report legib#lty. The reports and mit be nported in accordance with the syst:ms should be functioning, failure to copies that licensees are required to procedures described in 150.73 (b) (c). recognize the true nature of the event) or submit to the Commission undir the (d). (s). and (g) of this chapter and must include the information requiredin a procedural error; provisions of this section must be of paragraph (b)of this section. Events (ill Whether the error was contrary to sufficient quality to permit legible reported in accordance with 150.73 of an approved procedure, was a direct repruduction and micrographic this chapter need not be reported by a result of an error in an approved rocening a Mp er paugnpW procedure, or was associated with an (f) Demptions. Upon written request this sectim. activity or task that was not covered by from alicenses including adequate

4. In l 20A03, the introductory text of g y g p g,,,

,,,f) Any unusual cha'racteristics of the jusuncauon or at the inidade of the. persgraphs (a) and (b) is revised, and (// NRC staff, the NRC Executive Director paragraph [d)is revised to read as' worklocation(e.g heat. noise) that f r Operati ne may, by a letter to the follows: directly contributed to the erron and Heensee, grant exempdons to the 'lir/De type of personnelinvolved (i.e contractor personnel, utility. r* porting requirements under this, i 20A03 NoMcations cynchts, section. (a)Immediate notificollon. Each licensed operator, utility nonlicensed cperator. other utility personnel). (g) Reportable occurrences.The licensee shallImmediately report any (K) Automatically and manually requirements contained in this section events involving byproduct, source, or initiated safety system responses. replace all existing requirements for special nuclear material possessed by (L)The manufacturer and model licensees to report ** Reportable the Ifeensee that may have caused or threatens to cause: number (or other identification) of each Occurrences" as defined in Individual component that failed during the event. plant Techrdcal Specifications. (3) An assessment of the safety %e following addidonal admda (b) 7kenty-four hour notificotion. cach licmsee shall within 24 hours of - consequences and implications of the are also made to parts 20 and 50 of the event.This assessment must include the regulations in this chapter, discos ery of the event report any event availability of other systems or invcMng i censed material possessed

I 33860 Fed:ral Regist:r / Vol. 48, No.'144 / Tuesday, July 26,1983/ Rules and Regulations by the licensee that may have caused or (iv) Corrective steps taken or planned (i) Licensees that have an installed threatens to cause: to prevent a recurrence. Emergency Notification System shall make the initial notification to the NRC (d) Reports made by licensees in (c)(1) In addition to any notification Operations Center in accordance with response to th requ'rements of this required by i 20.403 of this part, each (50.72 of this part. section must c.nade as follows: licensee shall make a report in writing of (ii) All other licensees shall make the (1)1.icensees that have an installed levels of radiation or releases of initial notification by telephone to the Emergency Notification System shall radioactive material in excess oflimits Administrator of the appropriate NRC make the reports required by paragraphs spe'cified by 40 CFR Part 190, Regional Office listed in Appendix D, [a] and (b) of this section to the NRC " Environmental Radiation Protection Part 20, of this chapter. Operations Center in accordance with Standards for Nuclear Power (7) Written repons. Holders of an 150.72 of this chapter. Operations " or in excess of license operating licenss~for a nuclear power-(2) All other licensees shall make the conditions related to compliance with 40 plant shall subnJt a written report to the reports required by paragraphs (a) and CFR Part 190. Commission concerning the incidents - (b) of this section by telephone and by (2) Each report submitted under included in paragraphs (c) (1) and (2) of telegram, mailgram, or facsimile to the paragraph (c)(1) of this section must this section in accordance with the Administrator of the appropriate NRC describe: procedures described in 15053 (b),(c), Regional Office listed in Appendix D of (i) The extent of exposure of (d),(e), and (g) of this part. Incidents this part. Individuals to radiation or to radioactive reported in accordance with 150.73 of material; this part need not also be reported under

5. In i 20.405, paragraphs (g) and (c) are revised, and new palagraphs (d) and (ii) Levela of radiation and paragraphs (c)(1) or (2) of this section. -

concen' rations of radioactive material Dated at Washington. D.C. this 20th day of (e) are added to read as follows: involved: July loss. ~ j 20.405 Reports of overeuposures and (iii) The cause of the exposure, levels, For the Nuclear Regulatory Commiss' ion. excesaNe levels and concentrwoons. or concentrations; and - - Samuel [ CM. ~ (a)(1)In addition to any notification (iv) Corrective steps taken or planned y j required by 120.403 of this part, each to assure against a recurrence, including licensee shall make a report in writing the schedule for achjeving conformance [M Dw. amu ham ahl concerning any one of the following with 40 CFR Part 190 and with amas coot ru+m types of incidents within 30 days of its associated license conditions. occurrence:. (d) For holders of an operating license. (i) Each exposurt of an individual to for a nuclear power plant the incidents radiation in excess of the applicable included in paragraphs (a) or (c) of this lirnits in il 20.101 or 20.104(a) of this section must be reported in accordance part, or the license: with the rocedures describe ~d in (ii) Each exposure ef an individual to paragrap a 5053 (b),(c), (d),(e), and (g) radioactive materialla sxcess of the f this chapter and must also include the applicable limits in il 20.103(a)(1), inf nnation required by paragraphs (a) 20.103(a)(2), or 20.104(b) of this part, or and (c) of this section. Incidents in the license: reported in accordance with i 50.73 of (iii) Levels of radiation or this chapter need not be reported by a concentrations of radioactive materialin duplicate report under paragraphs (a) or secuon. a restricted area in excess of any other (e) All therlicensees who make applicable limit in the license: rep ris under paragraphs (a) or (c) of (iv) Any incident for which this section shall within 30 days after notification is required by i 20,403 of learning i the overexposure or this pan; or excessive level or concentration, make a (v) Leveis of radiation or report in writing to the U.S. Nuclear concentrations of radioactive. material Regulatory Commission. Document (whether or not involving excessive Control Desk. Washington. D.C. 20555, exposure of anyindividual)in an with a copy to the appropriate NRC t unrestricted area in excess of ten times Regional Office listed in Appendix D of any applicable limit set forth in this part this part. or in the license. (2) Each report required under PART 50--DOMESTIC LICENSING OF paragraph (a)(1) of this section must PRODUCTION AND UTILIZATION desenbe the extent of exposure of FACILITIES individuals to radiation or to radioactive material, including:

6. In i 50.36, new paragraphs (c)(6) and (7) are added to read as follows:

(i) Estimates of each individual's exposure as required by paragrsph (b) 150.36 Yechnical specencations. of this section: (ii) levels of radiation and (c) * * *, concentrations of radioactive material (6)Initia/ Notification. Reports made involved; to the Commission by licensees in (iii)The cause of the exposure, levels response to the requirements of this . or concentrations; and section must be made as follows:

[INTENTIONALLYBLANK] Second Draft, NUREG-1022, Rev. I

i 4 i 4 APPENDIX F 1992 REVISION TO 10 CFR 50.72 AND 50.73 INCLUDING ) STATEMENT OF CONSIDERATIONS .l ~ \\ a d i Published in the Federal Register on September 9,1992 September 10, 1992 (Vol.57,No.176,pages41378-41381) i NOTE: 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. l Its purpose here is to present the Statement of Considerations, which-explains some of the reporting requirements of the sections. ) 1 b l l i A f Second Draft, NUREG-1022, Rev. I

l 4578 FederalRegister /.Jol57db r170.7/; Thursday,> September:10.31902 /eRulesmWMesulations,.n..i.. certified to Ohm. In a letter dated For the Nuclear Regulatory Commission.. 50.72 and 10 CFR 50J3--Clarification of, ' August 14.1992 that by unanimous vote Samu 41.rm NRC Syatems and Guidelines For. the Commission had overridden the secmory cf rbe commi,,/en. Reporting." Followtna resolution of.. OhG's disapproval of the information IFR Doc. 92-21754 Filed G4-02: e.45 ami public comments, the NUREG will be collection request associated with this issued in the final form.%e NUREG samo cooc mm rule. will contain improved guidance for On August 21.1992. Oh2 assigned the event repomng. following new control number 31% 10 CFR Part 50 NRCs reviews of operating 0171. effective until August 31,1995. RIN 3150-AE12 experience and the patterns oflicensees* his new control number is only reporting of operating events since 1984 applicable to the sections in to CFR part M!nor Modifications to Nuclear Power have indmated that reports on some of l 35 amended by this rule. information Reactor Event Reporting these events are not necessary for the collection authonty for all other sections Requirements NRC to perform its safety mission and of 10 CR part 35 remains under the that continued reporung of these events existing general control number: 31% Aor.Ncy:Ntidear Regulatory would not contribute usefulinformadon 0010. Cmnminion. to the operating reactor events. A D N: Final rule-List of Subjects in 10 CFR Part 35 database. Additionally these. SUMS &Afty:ne Nuclear Regulatory unnecessary reports would have Dyproduct material. Criminal penalty. Commission (NRC) has amended its conunued to consume both the 3. Drugs. Health facilities. Health regulations to make minor modificadons licensees' and the NRCs resources that .p.rofessions. Incorporation by reference, to the current nuclear power reactor - could be better applied elsewhere.De, Medical devices. Nudear materials, event reporting requirements. The final NRC has determined that certain types Occupational safety.and health- . rule applies to all nuclear power reactor f events, primardy,6ose invoidag.. Radiation protection. Reporting and licensees and deletes reporting Invalid engineered safety fea1ure (ESF). recordkeeptng requirements. requirements for some events that have actuations are oflittle or.no, safety Text of Final Regulations been determined to be of little or no sWicance. safety significance.ne final mie* Valid ESF actuauons are those. reduces the industry's reporting burden, actuations that result from " valid.n -. l-For the reesons set out in the l preamble and under the authority of the and the NRCs respo'tse burden in event signals" or from intentional manual. Atomic Energy Act of1954, as amended, revltrw and assessmet.t.' initfadon, unless it is partlof h *., .H as amended, and 5 U.S.C 552 and 553.3 EnECnVE m OctA.111992. preplanned test. Valid signals are those f the Energy Reorganization Act of 1974. a ~ n - ~ s!gnals that are initiated itt, response to J the'NRC is adopting the following > "" " *E".8 " "8"" M N ACE actual plant conditiotis of.patameters'.o OW h [*, PART 35-MEDICAL USE OF '., Ev I on f ra na Dat LS Nuclear Regulatory. Commission... . g

- N j

DYPRODUCTMATEHtA1. n,,DCg)S55.Llephone(361).' those en d teria Il 1.%e authority citation for part 35 suPft.aw ETAJ5 inh:ORsdAh0N: include actuation's that are not the result

continues to read in part as follows:

of valid signals and are not intentional Back "und 3 manual actuations. lavalid actuations Authorttr Sees.161. e4 Stat. 94s. ss mended (42 LLS.C. not1: sec. 201. 88 Stat. De Commission is issuing a final rule include instances where instrument. 1242. se amended (42 UAC. 5641) * * *. that amends the riuclear power reactor

2. In i 35A paragraph (b)is revised event reporting requirements contained. drift, spurious signals, b6 man error, or and paragraph (d)is added to read as in to Cm 50.72. "Immediate Notification other invalid signals caused. actuation of the ESF (e.g, jarring a cabinet, an error follows:

Requirements for Operating Nuclear in use of jumpers of liftedleads, an error. Power Reactors." and to CFR 5033. in actuation of switches oc controls. I35.s informatkmcosectkm "!Jcensee Event Report Syatern." ne equipment failure, or redio frequency. requirementa: ONS approvst final rule is issued as part of the interference) Commission's ongoing activities t (b) ne approved information improve its regulations. Specifically, this NRCs evaluation of both the reported collection requirements contained in this final rule amends to CFR 5052 (b)(2)(li). events since January 1964, when the - part appear in il 35.12/35.13; 35.14 and 10 CFR 50.73 (a)(2)(lv). On June 20' e isting rules first becama effectiv' e. and h a dived during the' Event. ' 35.21, 35.22. 35.23.'35.27. 35.29. 35.31. 1992 (57 FR 30642), the Commission 35.50.35.51.35.53,35.59,35.60,3541. lasued a pmposed rule requesting public Reporting Workshops conducted in Fall - 3570, 35.80, 35.92. 35J04; 3525, 35.310. comments on 6ese amendments.. ~ of1990 identified heeded vements. 35.315.~35.404. 35.406,35.41'O 35.415 Over the past several years, the NRC in the rules.%e NRC dd ealhab' 35.006 35.810,35415,35.630. 35.632, has increased its attention to event i M id M M M M 35.834. 35.636, 35.64L 35M3. 35.645, and Mporung lasus to ensun undormity, mM a Mht Ms 35M7' consistency, and completenes:in including the systems, subsystems or re rtirypin Sepfember 1971, the NRCs c mponents (i.e an tavalid actuation.. D ce forAnalysis and Evaluation of isolation, or realign:hent'of only' the (d) OMB has soignad control number Operadonal Data (AEOD) issued for nactg water clean up(RWCU) system. 315060171 for the information collection comment a draft NUREC-1022. Revision ~....A requirements contained in iI 35.32 and. 1,i " Event Reporting Systems to CFR Ncl**' k*rdaC" bie for ing.caos oc. 3M Washins 35.33. 25s. A coPr 68 al*e avaat ~ copyms for a fee et the NRC Pubuc Doosment Dated at Rockville. Maryland. this 31 day Prees4nske copy snay be retoested by wnung to Room. *t20 L Sciet. NW414wer Leve4.. cf Septemt>er 1992. theDismbeuce and Med Services Section. UA WasMastoa. DC 2o565. 1

. Federal RagisterdWalh54No, soy /4Annkiswa64tctulic@oA19aMIMWJWNftdd=913Wgg .f y. the control room emergency Tentilation.[ e. ventilation system, auxiliary buildingggninul,ecdadesihr (CREV) system, the reactor building ~ ventilation system.ortheirequivalent4 events fromreportingre..y.,,.M;-N ventilationsystem, the fuel building ventilation systemsk ne' actuatiun of. (1).The firstcategory excludes events - ventilation system, or the auxillary the standby gas treatment system ' in which an in. valid ESFor,RPS.. building ventilation system. or their. following an invalid actuaticiof the. actuation occurs when the system is equivalent ventilation systems) are of : reactor building ventilation system is already properly removed from service. little or no safety significance. However,, also exempted from reporting. In if all requirements of plant procedures these events are currently reportable addition, the final rule excludes invalid for removing equipment from service - under 10 CR 50.72 (b)(2)(li) and to CFR actuations of these ESFs (or their have been met.hisincludes required 50.73 (a)(2)(iv). equivalent systems) trom signals that clearance documentation, equipment The final rules for the current event onginated from non-ESF ciremtry. and control board tagging /and properly reporting regulations.10 CFR 50.72 and ., However, invalid actuations of other positioned valv,es and power supply., 10 CR 50.73 (48 FR 39039; August 29 ESFs would continue to be reportable. b eakus. 1983, and 48 FR 33850:]uly 26,1883,

*mpj g

(2) The second category excludes - respectively), stated that ESF systems; - g,alens ~ ! d. . events in which an invalid ESF br.RPS : includios the reactor protection system" actuation occurs after the safety.~". v. i lad (RPS), are provided to mitigate.the [,"t af7f c cling tenis, ma eam. function bas alreadybeencomplet' d.T e , ' consequences of a significant event. J "#'"3 (e.g, an invalid containmentisolation. :/ Therefore, ESFs should(1) work-

[*"' g '

',I* "" Ignal while the containment isola'tloir? 9 p g properly when called upon and (21 should not be challenged frequently or residual heat removg system isogations valves are already closed, oran invalld e: unnecessarily, ne Statements of. . (orsystems design'ated any other ~ gg ,.,gg Consideration for these final rules also names stg o fu the , gg

4.,

b{3) g,.~ hird category excludes events 1 ' t stated that operation of an ESF as partw function s ,rto ese systems Md ESF WMUM ~ cf a pre-planned operational prdcedure. their equivalentsk are suu report 4W.E occurs that involes only'ar limited or test need not be reported.The , an invalid ESF actuation reveals s'.. C:mmission noted that ESF actuations. defect in thesystem so that the system

g g.

4.b ' ' ~ failed or would fall to perfonn its. including reactor trips, are frequently . intended function.-the event contin, es to,. RWCU Vyjfem or sifyof tb.s, lo'lfowi.n.g /

casociated with significant plant :

u y g ~,. transients and are indicative of events..... be reportable under other reqmrements that are of safety sieninne. At that of to CR 50.72 and10 CFR 5&73.If a reactor buil&,g venulatlan* system',' k, ng yen on sys$ saxiBar i bundingvendanmam%ar%p,,, time, the Commlulon also required all condition or deficiency has (1) an. ESF'actuations, including the RPS. adverse impact on safety-related r, actuations,whethermanualer...., equipment and consequently on the., !. mula .g. . sitomatic, valid orinvalid-except as - ability to shutdown the reactor and ' N*- - ESFs not'sp y aQfm. noted..to be reported to the NRC by-maintain it in a safe shutdown x ' '/.M. emergency Mcmung systanc telephone within 4 hours of occurrence condition,(2) has a petentialfor. - potential exposure toTlaat personnel or. Iso ation valy,a,ctua anonser followed by a writtenIMaae Event 4 significant radiological'reles s'd dr s closuree thaliaffect . Report (LER) within 30 days of the, .coohng systeinh h b,y,,,, lacident.%1s requirement on. timeliness the general public, or (3)'would. r eSaedal s.upport systems, e,% : . of reporting remains unchanged.. - compromise contral room habitability,~ e utainment spray actuation; residtial -, The reported informationis used by the event / discovery continues to bea thi NRC in confinnation of the lic'ensing. reportable,. heat removal. system isolations, or their bases, identification of precursors to Invalid ESP sctuatleris tlist are s a se den of. excluded by this final rule, bu't ocS as - ,,, centinue't'o terErhired to p lessons, review of management con'rol - a part of a reportable event, continue to systems, and licensee performance be described as part of the reportable submit LERs if a deficiencforconditidu t associated with any of the invelid ESF.

emt.nese amendments are not actuations of the RWCU of the CREV,

assessment.. intended to preclude submittal of a Discussion complete, accurate, and thorough systems (or other equivalent ventilation. The NRC has determined that some description of an event that is otherwise systems) satisfies any bility criteria under i 50.72 , ( th events that involve only invalid ESF reportable under 10 CFR 50.72 or 10 CFR actuations are oflittle or no safety sa73.The Commiselon relaxed only the - Impact of the Aad- ___ en the " r-significance. Hewever, not all invalid

elected event reporting requirements.

Industry and Government Resources . ESF actuations are be exempted from, specified in this final rule-Relaxing the requirein'ent for reporting reporting through this r ne Licensees are still redited under 10 of certain types of ESFactuations .retaxations la event reporting... Cm part 30. appendix IL " Quality reduces the industry's burden. requirements contained in the final rule Assurance Criteria for N6 clear Power and the NRC's risponse en. RIs apply only to a narrow, limited set of plants and Fuel Reprocessing Plants." to reduction is consistent with the - specifically defined invalid ESF. address corrective actions for events or objectives and the r@;;m.ta of the Metuations.These events include invalid. conditions that are adystsa to quality. Paperwork Reduction Act.These ' actuation. isolation,orreaHynment of a whether the event is reportable or not. -, amendm'ents have noimpact on the ~ ,Jimited set of ESFs including systems,. In addition. mNi*ing ESF a ctuations :. NRG's ability to fulfilllis mission to -, ~ subsystems. or components (i.e., an (such as RWCU !solat!ans) to reduce ensure public health and safety.becauise , invalid actuation. Isolation, or.. operational radiation exposures - the deletedreportability requirements ' realignment of only.the.RWCU.uystem, associated with the investigation and havelittle nr no safety sieninum. or theCREV system.reactoe bWing recovery'from the petuations, are It is estimated that thichanges.talhe '. ~ ventilation system,fuelbuilding - consistent with ALARA requirements. e'xisting rules will resull ifabout.150 for -.

0bWb2&hh$$Yfb $$YYESS $Y YY$W-asm~ %-10'percMt)feMrditbMviEf*$ ' theletoticerns abo'et eliminating the average 50 hours per licenace response. Reports each year. Similar reddctiotisM selected event reporting raquirements, including the time required reviewing areexpectedinthe numbertrfprompt < These commenters believe that the instructions, searching existing data event notifications veportable tmder 10 elimination of these event reporting sources gathermg and snaintaming the. CFR 5052.Some respondents.intheir ' requiretnents may adversely affect the data needed, andreviewing the comments on the proposed rule. dated NRC's information database and collection ofinformation. Send June 20.1992. submitted an estimate of ultimately affect the agency's ability to comments regarding the estimated a ppro ximately 15 percent red uction in carry out its mission to protect public burden reduccon or any other acect of j their reporting trutden. health and safety. For many years the this collection of information. including Summary of Comments NRC staff has been systemat2cally suggestions for reducing this burden, to reviewing information obtained from the Information and Records The NRC received 19 comments-2 Licensee Event Reports. These Management Branch (MNBB-m4). U.S. from individuals. 3 from industry-assessments of reactor operstional Nuclear Regulatory Ccmmission. supported organw%and 14 from expenence have indaded data on the Washington. DC20557c and b the Desk utilities. Except ice two responden ts. all types of events toduded in the hree Officer. Office ofinfeinstion and 1 commenters wek:ctned the :- categories that the NRC is deleting from Regulatory Affafts. NEOMoto. (3150-4 j Cornminaion's e%rts to nduce dwe reporting. The staffs reviews and 0011 and St$0-0104). Office of licensee burden and tosave the as seaaments of osarly 1000 reactor-Management and Budget. Washingto't. 1 agency's resources in event review and yeare of operational experience have DC 20503. processing.& utdities and the. identified essentially no eafety industry-suppodedorganizations significance associated with the type of Regulatory Analysis 3 1 expressed their desire for a broader events included in the aforementioned relaxation to f vinda all invalid ESF. three categories.h r%mmissionhas h ConnisAonhas pmpend a actuaticas troot reporting.w. a. . reviewed the scope of thena ngulatmyanalysisonW finalrule. j Othercommenta6nuntherespondents amendments.andon the basis of the h analysis examines the mer and concerned the fonowing:darificatbn of etaffa assessment of the past reactor benefits of the attematives considered the definitica ni.' invalid

actsations:,

operationalexperience.has by the Commission.The nnalysis ts examples of awntsbeing exempted 'z.4 subsequently concluded with a available forinspection inthe NRC from reporting: consideration of piant-reasnnaMa confidence that relaxation Public DocumentRoom.2120 LStreet, from reporting of events in the three NW. Lower Level. Washington. DC apecific situations:mmption from .. s reporting of thasctuationof thestandby; categories does not afIect theagency's 20555. Single copies e(the analysis may gea trastment system following an ability to protect public health and be obtained imm: Ra iTripath!. Office invalid actuation of the raa* building safety. for Analysis and E ustion of ventilattu systesa:aad poasibly Based on the laput froin tire utilities, Operational Data.U.S. Nuclear m extending thh ofinvalid. these amendments will reduce the Regulatory Commisalon. Washington. I actuationslisolations of RWCU from. industry's reporting burden by about 15 DC 20555.Telsphons (301) 424435. re to ivinA= 1 hose of the. - percent.The eatimated savings of the - chemi and volume controlsystemin NRC's response burden in event review Regulatory FlaxibilityCertifscation a pressurizedwaterreactor.b. and assessment is about 5-10 percent. In accordance with the Regulatory i Statement of Coc41derations for this I finalrule addresses most of these Envinmmentahmpact Categas. cal Flexibility Act of 1980 (51,LS.C. 605 (B)). the Commission certifles that this. rule comma. Othatissuesanddanfications. Exclus4on does not have a aqmificant economic concerning event nportability will be. - % NRC has determined that this impact on a substandal number of small addressed in NUREG-1022. Revision 1, final rulsis the type of action described entities The final rule affects only the However, it is cot practicallo address a in categorical exclusions to CFR 61.22 eve'nt reporting requirements for plant. specific untion unlesa it relates (c)(3)(li) and (MI). Wrefore, neither an - operationalnuclear power plants.W to a genenc cot ern. environmental impact sta tement nor an companjes that own these plants do not & Commision stresses that only environmental assessment has been fall within the scope of the definition of certain specific invalid ESF actuations prepared for this final rule. "small entitles" set forth in the "[ [p*[ Paperwork Reduction Act S'atement e a e b ty Act or e Small NUREG-1022. Revision 1 will contain This final rule amends information specific examples and additional coUection requirements that are subject regulations issued by the Small Business guidance on events which are presently to the Paperwork Reduction Act of 1980 Administration Act in 13 CFR part 121. i reportable as well as those which are (44 U.S.C. 3501 et seq.). These Backfit Analysis bemg exempted from reporting through amendments were approved by the a these amendments. In the future, the Office of Management and Budget As required by 10 CFR 50.109. the. Commission will give due consideration approvalnumbers 3150 0011 and 3150-Commission has completed an to other proposed relaxations from 0104. assessment of the need for Beckfit event reporting after the NRC staff has Because the rule will relax existing Analysis for this Tmal rule.ne had an opportunity to reassess the data reporting requirements, public reporting proposed amendments include needs of the agency andperfonned burden ofinformation La expected to be relaxations of certain e.xistin; safety assessments to Justify initiating a ' reducedelt is estimated that about 150 f*quimments on repeting of nnatum ~ separate general rulemaking.Until such fewer Licensee Event Reports (NRC. to the NRC %ese changes neither. time, all events not.spscifically Form 368) and a similarly reduced impose additional reporting exempted in these amendments continue. number of prompt event notifications. ' mquirements not requits modifications to be reportable.. ' made pursuant to 10 CFR 5072, will be. to the facilities or.thalt!! censes. W two respondents who opposed ' required eoch year.The resulting Accordingly. the NRChas concluded the proposed amendments expressed reduction in burdenis estimated to that this final rule does not constitute a

p e.--- LFederal.'Registerd%1e57bNr 917tF/aThursddypBeptembetW10.3992PRtdssfatnd$4gi21kR$s,L9 B8UJ,f - backfit and, thus, a backfit^ analysis is -

2. In l 50.7L paragraph (b)(2)(li) is' For the Nuclear Regulatoryprunndulon.},,,

not required. V' revised to read aa follows: - Jama R Taylor. r' ~ ,f' [ !Jat cf Subjects in 10 CFR Part 50 g 50.72 immee. ate notmestion ,,y, Antitrust. Classified information. P** su.uno coot rse m.a Criminst penalty. Fire prevention. incorporetion by reference. Intergovernmental relations. Nuclear (b) Non-emergency Events. * *

FEDERAL RESERVE SYSTEM power plants and reactors. Radiation (2) Four-bour reports.

protection. Reactor siting criteria. (ii) Any event or condition that results 12 CFR Part 225 Reporting and recordkeeping, in a manualor automatic actuation of ) For the reasons set out in the. any enginected safety feature (ESF). (Regulat$on Y; Docket No.R-0704) preamble and under the authority of the including the reactor protection system Atomic Energy Act of 1904 as amended. (RPS), except whem the Energy Reorgamzation Act of 1974. (A) he actuation results from and is Bank Holding Companies and Change as amended, and 5 U.S.C. 552 and 553 part of a pre-planned sequence during the Commission is adopting the testing.or reactor ope ation:. in Bek Control . fillowing amendments to to CFR part (B) ne actuation is invalid and: AGENCv: Board of Covernors of the 50. (f) Occurs while the system is-Federal Reserve System.,. properly removed from service: AcnoN:Pinal rule.. ~ PART 50- DOMESTIC (JCENSING'OF (2) Occurs after the safety function .PH000CT10N AND UTi!.!ZAT10N. has been already complaited:.or sussasAJt't:The Board is amending its FACIUTIES. (J) Involves only the following specific Reguletion Y to augment the list'of '1 ESPs or their equi &Qystems:- We nonMy l ' rev. The authority cita tion for Part 50 is (i) Reactor water clean-up system: - holding companias to e. ised 'to read as follows: Authorith. Seca.102.133.104l1o5, tet.182, - (ii) Control room emergency provision of fu!! service seemities .. ta3 tea, tae, es Stat. s:e. on.ssa, sea. s53. ventilation system: brokerage under certain conditions; and' 954. 955. 95e. as amended, s' c. 234. as Stat; (iii) Reactor building ventilation - the provision of financialadvisory e 1244. as sawnded 142 UAC 2132. 2133. 2134. system.. services under certain conditions.De 2135.2201.2222.2233.22se.22as,22a2h a ca.. (iv) Fuel building vent $ation system: Board has by orderpreviously pp ed 201.as amended.202.20tL 88 StaL1242 ae or >l' tbeae activities. Applications ' amended,1244.124a (42 UAC 5641 satz. - (v) Auxiliary'buildinhen....tilation holding companies to engage la.. 5446). activities included on the Regulation.Y - system, ' section 27 etso innd underPub.E os. ". M< a "..2 V. ' list ' f]$ermissible nonbanidag activities 7 ' o . cot.mc.1a s2 Stat. 2sst (42 UAC se51). l S. x t ~ ms.beprocessed by th4Reesrvehnks ; l Sectico Eato also honed under secs. t01:185'

3. In I 60.73, paragraph {a)(2)(iv) is..

yer". ted p"t==g,aish*Dt.,tT;', "p i C Stat. ** ess."as amended (42 UAC rist; revised to read as follows:a 223sh eec.102. Puth L et-too, as Stat.ss3 (42-delegaf authority..,.~.u.7, .o g . UAG 4332).Secekme m13. ms4(ddicand. 8 60.73. Ucensee event'r port system. EFFECTiva cATE:Siptem@r.10.1982. ". ' . mtas also neeued under sec tos, ce Stat. eso, (a) Reportable events.3,,.*.*, Fon runTHEn percasdAMON CostfACfr. [' se=== Mad (42 USC 2138). Sections 50.23. (2)ne ucensee shallieport t * * ' Sectt G. Alvarez. Associate Central-YggN,h3"38

jd", C' '

(iv) Any eventarcondition that Counsel (202/452-4583).or %omas M.. 3 60.33a. so35s. and Appendi.x Q also issued tesulted in a snanual or automatic Corsi. Senior Attorney (202/453,1275). : under sec.102. Pub. L st-tso 83 Stat. 853 (42 actuation of any engineered safety - Legal Division; For the hearing impaired UAC 4312). Sections m34 and 60.54 also feature (ESP) including.the reactor; only. Telecommunications Device for leeued under sec. 204, se Stat.1245 (42 UAC protection system (RPS). except whent the Deaf (TDD). Dorothea noenpson se44% Sectiona nasa, mat. and san 2 also (A) ne actuation resulted from and (202/452-3544).' sMiso W88 part of a pre-planned sequence SUPet.EstwrARY INFCas4AD0st'. under during testing or reactor operation: (B) The actuatior: was invalid and:

Background

el t und r e sa 864, as amer.ded (42 UAC 2234). Appendix F - (f) Occurred while the ayatem was The Bank Holding Company Act of also luued ander sec.1s7. ea Stat. oss (42 properly removed from service: 1956, as amended (the "BHC Act"), l UAC 22n). (2) Occurred after the' safety function generally prohibits a bank holding For the purycees of uc.123. 68 Stat. osa, as had been already coinpleted; or company from engaging}in Mah amended (42 USC 2273h i I so.s. sa4e(s) (3) Involved only the following activities'or si: quiring vsting'securitici asEe U1"C specmc ESFs,or their equivalent of an'y compainythat is not a bank. ' t. ( 2201(blh i I 5as. Sa'11al. so.to( He), sos 4(e) systems: ~, Section 4(c)(8) of the BHC Ai:t)tovides - and (e), so.44(eHei. m4e(s) and (b). sa47(b). (/) Reactor water clean.up system;. an exception to this prohibiti6u where sa4s(et M (d), and (et mes(sh so.54(a), (1), Ui) Control room emergency the Board determines after notfoe and-010) (1Hnk (pl. (q),(t). Iv1 and ty), so.ss(r),. ventilation system: opportunity for hearing that the- - saaseM (cHeh (s). and(h). mso(c). (iii) Reactor building ventilation activities being conductedsiire "so saso(ak So s2(bk so.64(b) 5045. and SoJo(a) system: closely related to benWs'or managing gg**cges Stat. (iv) Fuel bizilding ventilation system:. or controlling banks'as to be'a' proper YsN(b v) Auxiliary bul!Ill'nghentilation'. Iho a e t a :( saro(et sa711aHe) and M so.72te), so.73(a) system. determination by order in an individual and (b). 6a74. 50.78 and 50.90 are tasued. C C case or by regulation. ' under4ee, tetoies Stat. 850. se amended (42 - Deted at Rockville. MD; thle 27th day of The Board's.Regulationy (12 CFR part UAC 220tto)). August.1992. 225) sets forth a list of nonbankihg

~ NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSIOrd

1. REPORT WUMBER (3-89)

(Assi Supp.gned tPJ NRC Add Vol.. NRCu tio2. , Rev.. and Addendum Num-32oi,3:o2 BIBLIOGRAPHIC DATA SHEET "" ""Y 1 (see instructions on the rev.rse) NUREG-1022, Rev.1 - SeCond Draft

8. TITLE AND SUBTITLE

3. DATE REPORT PUSUSHED Event Reporting Guidelines j

Monm YEAR 10 CFR 50.72 and 50.73 Second Draft for Comment February 1994-

4. FIN OR GRANT NUMBER

5. AUTHOR (S)

6. TYPE OF REPORT D. E Allison, M. R. Harper, S. Israel, W. R. Jones, J. B. MacKinnon, S. Sandin -

Regulatory

7. PERIOD COVERED (inclusive Dates)

G. PERFORMING ORGANIZATION - NAME AND ADDRESS (tf NRC. provide Division. Office or Region. U.S. Nuclear Regulatory Commission, arid maeling address; if contractor, provide name and malling address.) Office for Analysis and Evaluation of Operational Data U.S. Nuclear Regulatory Commission j Washington, DC 20555-0001 1 l

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (tf NRC, type "Same as above"; 11 contractor, provide NRC Division. Office or Region, U.S. Nuclear Regulatory Commission, and malling address.)

l Same as above.

10. SUPPLEMENTARY NOTES

11. ADSTRACT (200 words or less)

Revision 1 to NUREG-1022 clarifies the immediate notification' requirements of Title 10 of the Code ofFederal Regulations, Ibrt 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 1 document was noticed for public comment in the Federal Register on October 7,1991 (56 FR 50598). 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.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.)

'13. AVAILABLITY STATEMENT Unlimited

14. SECURlW CLASSIFICADON Emergency Notification System (This Page)

Event Report ~ Immediate Notification Unclassified i-Licensee Event Report . (This Rsport) Notification Unclassified Report

15. NUMBER OF PAGES

16. PRICE NRC FORM 335 (249)

l l l Printed on recycled paper Federal Recycling Program

UNITED STATES FIRST CLASS MAIL NUCLEAR REGULATORY COMMISSION POSTAGE AND FEES PAID WASHINGTON, D.C. 20555-0001 uSNRc PERMIT NO. G-67 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE. $300 3gc11911"'

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