ML20133E829

A Prioritization of Generic Safety Issues

ML20133E829
Person / Time
Issue date:	12/31/1996
From:	Emrit R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:
References
NUREG-0933, NUREG-0933-S21, NUREG-933, NUREG-933-S21, NUDOCS 9701130173
Download: ML20133E829 (174)

Text

_ _ _ _ _

A ttrug UNITED STATES g

NUCLEAR REGULATORY COMMISSION f

WASHINGTON, D.C. 20806 4 001 Os %,***.*,o/

I DECEMBER 1996

~

SUPPLEMENT 21 TO NUREG-0933

1. A PRIORITIZATION OF GENERIC SAFETY ISSVES" REVISION INSERTION INSTRUCTIONS i

Remove Insert

==

Introduction:==

pp. 29 to 62, Rev. 20 pp. 29 to 62, Rev. 21 i

Section 3:

pp. 3.15-1 to 9, Rev. 2 pp. 3.15-1 to 9, Rev. 3 pp. 3.83-1 to 2, Rev. 1 pp. 3.83-1 to 2, Rev. 2 Appendix A pp. A-1 to A-7 pp. A.A-1 to 6, Rev. 1 Appendix B pp. A-9 to 20, Rev. 12 pp. A.B-1 to 12, Rev 12,

Appendix C pp. A-21 to 23 pp. A.C-1 to 4, Rev. 1 pp. A.D.0-1 Appendix D l

pp. A.D.001-1 to 2 pp. A.D.002-1 to 2

]

pp. A.D-003-1 to 2 pp. A.D.004-1 to 3 pp. A.D.005-1 pp. A.D.006-1 to 2 pp. A.D.007-1 to 3 Appendix E pp. A.E-1 to 80 I

J s

V

&O 9701130173 961231 s

PDR NUREG 0933 R PDR

_ _. _ _. _ _.. ~_.. _ _

.m_

. _.__~,. - -_..

._.m_

O cn Nw Neon TABLE II LISTING OF ALL TMI ACTION PLAN ITEMS. TASK ACTION PLAN ITEMS, NEW GENERIC 1551K5. Als itslAN FACTORS ISSUES i

This table contains the priority designations for all issues 11sted in this report. For those issues found to be covered in other issues described in this document, the appropriate notations have been made in the Safety Priority Ranking colimi, e.g.. I.A.2.2 in the Safety Priority Ranking colism means that Item I.A.2.6(3) is covered in Itasi I.A.2.2. For those issues found to be covered in programs not described in this decisment, the notation (S) uns made in the Safety Priority Ranking colisen. For resolved issues that have resulted in new requirements for operating plants, the appropriate analtiple.at 1icenstng act1on nisaber is Itsted. The 11 censing action msubering system bears no relationshtp to the nisabering systems used for idontifytng the priorittred issues. An explanation of the classification and status of the issues is provided in the legend below.

Leeend NOTES: I - Possible Resolution Identified for Evaluation 2 - Resolution Available (Docismented in NutEG, NRC Memorandum, SER, or equivalent) l 3 - Resolution Resulted in either: (a) The Estabitshment of New Regulatory Requirements (By Rule. SRP Change, or equivalent) or (b) No New Requirements 4 - Issue to be Prioritized in the Future 5 - Issue that is not a Generic Safety Issue but should be Assigned Resources for Completion HIGH

- High Safety Priority IEDIUM

- Medium Safety Priority LOW

- Low Safety Priority DROP

- Issue Dropped as a Generic Issue EI

- Envirorumental Issue I

- Resolved TMI Action Plan Item with toplamentation of Resolution Mandated by INNtEG-0737 LI

- Licensing Issue l@A

- Multiplant Action NA

- Not Appitcable RI

- Regulatory Impact Issue 5

- Issue Covered in an NRC Program Outside the Scope of This "p ument t

USI

- Unresolved Safety Issue

,M i

c D

rv1 M

Q

^

s O

O 3

eW N

W

Table II (Continued)

Action Lead Office /

Safety Latest oas Plan Itan/

Priority Division /

Priority Latest Issuance NPA D Issue No.

Title Engineer Branch Ranking Revision Date No.

?.

c)

TMi ACTION PLAN ITEMS y

_0PERATING PERSONNEL I,A.1 Operattna Personnel and Staffina 1.A.1.1 Shift Technical Advisor NRR/DHFS/t08 I

2 12/31/86 F-01 I.A.I.2 Shift Supervisor Administrative Duties NRR/DHFS/LQti I

2 12/31/86 1.A.1.3 Shift Manning NRR/DHFS/LQB i

2 12/31/86 F-02 1.A.1.4 Long-Tem Upgrading Colmar RES/DF0/W8R NOTE 3(a) 2 12/31/86 W

Trainina and Dualtftcations of Operatina Personnel I.A.2.1 Isumediate Upgrading of Operator and Senior Operator Training and Qualifications I.A.2.1(1)

Qualifications - Expertence NRR/DHFS/LQB I

5 12/31/87 F-03 I.A.2.l(2)

Training NRR/DHFS/LQ8 I

5 12/31/87 F-03 I.A.2.l(3)

Factitty Certification of Competence and Fitness of NRR/DHFS/LQB I

5 12/31/87 F-03 Applicants for Operator and Senior Operator Licenses I.A.2.2 Training and Qualtftcations of Operations Personnel Colmar NRR/DWS/LQB NOTE 3(b) 5 12/31/87 NA I.A.2.3 Administration of Training Programs NRR/DMS/LQB I

5 12/31/87 I.A.2.4 NRR Participation in Inspector Training Colmar MRR/DHFS/LQB LI (NOTE 3) 5 12/31/87 NA go I.A.2.5 Plant Drt11s Colmar NRR/DHFS/LQB NOTE 3(b) 5 12/31/87 NA I.A.2.6 Long-Tem Upgrading of Training and Qualifications I.A.2.6(1)

Revise Regulatory Guide 1.8 Colmar NRR/DHFT/WIS NOTE 3(a) 5 12/31/87 NA I.A.2.6(2)

Staff Revtem of NRR 80-117 Colmar NRR/DWS/LQ8 NOTE 3(b) 5 12/31/87 NA I.A.2.6(3)

Revise 10 CFR 55 Colmar NRR/DWS/LQB 1.A.2.2 5

12/31/87 NA I.A.2.6(4)

Operator Workshops Colmar NRR/DHFS/LQB NOTE 3(b) 5 12/31/87

~NA I.A.2.6(5)

Develop Inspection Procedures for Training Program Colmar NRR/DNFS/LQB NOTE 3(b) 5 12/31/87 NA I.A.2.6(6)

Nuclear Power Fundamentals Colmar NRR/DHFS/LQ8 DROP 5

12/31/87 NA I.A.2.7 Accreditation of Training Institutions Colmar NRR/DHFS/LQ8 NOTE 3(b) 5 12/31/87 NA I.A.3 Licensino and Reaualtftcation of Operattna Personnel I.A.3.1 Revise Scope of Criteria for Licenstng Examinations Emrtt NRR/DWS/tQB I

5 12/31/86 1.A.3.2 Operator Licensing Program Changes Enrit NRR/DHFS/0LB NOTE 3(b) 5 12/31/86 NA I.A.3.3 Requt w ts for Operator Fitness Colmar RES/DRA0/HFSB NOTE 3(b) 5 12/31/86 NA I.A.3.4 Licensing of Additional Operations Personnel Thatcher NRR/DHFS/LQB NOTE 3(b) 5 12/31/86 NA I.A.3.5 Establish Statenent of Understanding with INPO and DOE Thatcher NRR/DHFS/HFE8 LI (NOTE 3) 5 12/31/86 NA I.A.4 Simulator Use and Deve1coment 1.A.4.1 Initial Simulator !sprovement

[

2 c-I.A.4.l(1)

Short-Tem Study of Training $1mulators Thatcher NRR/DHFS/0LB NOTE 3(b) 5 06/30/88 NA y I.A.4.l(2)

Interim Chriges in Training Simulators Thatcher NRR/DHFS/DL8 NOTE 3(a) 5 06/30/88 7

o I.A.4.2 Long-Tem Training $1mulator Upgrade g I.A.4.2(1)

Research on Training Simulators Colmar NRR/DWT/HFIS NOTE 3(a) 5 06/30/88 0

m I.A.4.2(2)

Upgrade Training Simulator Standards Colmar RES/DF0/HFBR NOTE 3(a) 5 06/30/88 y 1.A.4.2(3)

Regueatory Guide on Training Simulators Colmar RES/DF0/HFBR NOTE 3(a) 5 06/30/88 U

!.A.4.2(4)

Review Simulators for Confomance to Criteria Colmar NRR/DLPQ/LOL8 NOTE 3(a) 5 06/30/88 O

O O

)

J d

N Table II (Continued)

Action Lead Office /

Safety Latest

$ Plan Item /

Priority Division /

Priority Latest issuance PFA g Issue No.

Title Engineer Branch Ranking Revision Date No.

?

I.A.4.3 Feasibility Study of Procurement of NRC Training Colmar RES/DAE/RSRB LI (NOTE 3) 5 06/30/88 NA i

Simulator I.A 4.4 Feasibility Study of NRC Engineering Co g uter Colmar RES/DAE/RSRB Li (NOTE 3) 5 06/30/88 NA Q

SUPP(RT PER$G41EL l

1,5.1 Pr -

- for Onerations j

I.8.1.1 Organization and F-,

-t Long-Tem Improvements I.8.1.1(1)

Prepare Draft Criteria Colmar NRR/DWT/HFIS NOTE 3(b) 3 12/31/86 NA 1.8.1.1(2)

Prepare Commission Paper Colmar NRR/DWT/HFIS NOTE 3(b) 3 12/31/86 NA I.8.1.1(3)

Issue Requirements for the Upgrading of Management and Colmar NRR/DWT/H T.

NOTE 3(b) 3 12/31/86 NA Technical Resources I.8.1.l(4)

Review Resp)nses to Detemine Acceptability Colmar NRR/DHFT/WIB NOTE 3(b) 3 12/31/86 NA I.8.1.l(S)

Review Implementation of the Upgrading Activities Colmar OIE/DQASIP/0RPS NOTE 3(b) 3 12/31/86 NA I.8.1.1(6)

Prepare Revisions to Regulatory Guides 1.33 and 1.8 Colmar NRR/DTS/LQB I.A.2.6(1),

3 12/31/86 NA 15 I.8.1.1(7)

Issue Regulatory Guides 1.33 and 1.8 Colmar NRR/DHFS/LQ8 I.A.2.6(1),

3 12/31/86 NA 15 I.8.1.2 Evaluation of Organization and Management Improvements of Near-Tem Operating License Applicants I.8.1.2(1)

Prepare Draft Criteria NPS/DWS/LQ8 NOTE 3(b) 3 12/31/86 NA I.8.1.2(2)

Review Near-Tem Operating License Facilities NiiR/DHFS/LQB NOTE 3(b) 3 12/31/86 NA 1.8.1.2(3)

Include Findings in the SER for Euh Near-Tem NRR/DL/0RA8 NOTE 3(b) 3 12/13/86 NA Operating License Facility 1.8.1.3 Loss of Safety Function 1.8.1.3(1)

Require Licensees to Place Plant in Safest Shutdour.

Sege RES LI (NOTE 3) 3 12/31/86 NA Cooling Following a Loss of Safety Function Due to Personnel Error 1.8.1.3(2)

Use Existing Enforcement Options to Accomplish Safest Sege RES LI (NOTE 3) 3 12/31/86 NA Shutdown Cooling i

I.8.1.3(3)

Use Non-Fiscal Approaches to Accomplish Safest Shutdown Sege RES LI (NOTE 3) 3 12/31/86 NA Cooling 1.0 2 Insoection of Oneratino Reacters I.8.2.1 Rev6se DIE Inspection Program I.8.2.l(1)

Verify the Adequacy of Management and Procedural

$sge 0!E/DQASIP/RCP8 LI (NOTE 3) 11/30/83 NA Controls and Staff Discipline I.8.2.l(2)

Verify that Systems Required to Be Operable Are Properly Sege OIE/DQASIP/RCPB L1 (NOTE 3) 11/30/83 NA Aligned I.8.2.l(3)

Follow-up on Completed Maintenance Work Orders to Sege DIE /DQASIP/RCP8 L1 (NOTE 3) 11/30/83 NA Assure Proper Testing and Return to Service

,M I.8.2.l(4)

Observe Surveillance Tests to Determine Whether Test Sege OIE/DQASIP/RCP8 LI (NOTE 3) 11/30/83 NA M

Instruments Are Properly Calibrated 7

E I.8.2.!(5)

Verify that Licensees Are Complying with Technical Sege OIE/DQASIP/RCPB LI (NOTE 3) 11/30/83 NA e

Specifications I.B.2.1(6)

Observe Routine Maintenance Sege ole /DQASIP/RCP8 LI (NOTE 3) 11/30/83 NA w

I.8.2.l(T)

Inspect Tere'nal Boards, Panels, and Instrument Racks Sege OIE/DQASIP/RCP8 LI (NOTE 3) 11/30/83 NA f

j W

for Unauthorszed Jumpers and Bypasses

?

t

Table II (Continued)

Actton Lead Office /

Safety Latest om Plan Item /

Priority Division /

Priority latest Issuance MPA N

Issue No.

Title Engineer Branch Ranking Revision Date No.

w?

e*

!.8.2.2 Resident Inspector at Operating Reactors Sege O!E/DQASIP/ORP8 LI (NOTE 3) 11/30/83 NA

!.8.2.3 Regional Evaluations Sege OIE/DQASIP/ORPB LI (NOTE 3) 11/30/83 NA I.B.2.4 Overview of Licensee Perfomance Sege O!E/DQASIP/ORPB LI (NOTE 3) 11/30/83 NA M

QPERATING PROCEDURES I.C.1 Short-Tem Accident Analysis and Procedures Revision 1.C.1(1)

Small Break LDCAs NRR I

3 12/31/86 I.C.1(2)

Ir. adequate Core Cooling NRR I

3 12/31/86 F-04 I.C.1(3)

Transients and Accidents NRR I

3 12/31/86 F-05 I.C.1(4)

Confimatory Analyses of Selected Transtents Riggs NRR/DS!/R$8 NOTE 3(b) 3 12/31/86 NA

!.C.2 Shift and Reitef Turnover Procedures NRR I

3 12/31/86 I.C.3 Shift Supervisor ResponsiblIities NRR I

3 12/31/66 I.C.4 Control Room Access NRR I

3 12/31/86 I.C.5 Procedures for Feectack of Operating Expertence to NRR/DL I

3 12/31/86 F-06 Plar.t Staff I.C.6 Procechres for Vertf tcation of Correct Perfomance of NRR/DL I

3 12/31/86 F-07 Operating Activities I.C.7 NSSS Vendor Review of Procedures NRR/DTS/PSR8 I

3 12/31/86 1.C.8 Pilot Monitoring of Selected Emergency Procecbres for NRR/DHFS/PSRB I

3 12/31/86 Near-Tem Operating License Appiteants w

to I.C.9 Long-Tem Program Plan for Upgrading of Procedures Riggs NRR/DHFS/PSRS NOTE 3(b) 3 12/31/86 NA Q

CONTROL ROOM DESIGN I.D.1 Control Room Design Reviews NRR/DL I

7 06/30/95 F-08 1.D.2 Plant Safety Parameter Display Console NRR/DL i

7 06/30/95 F-09

!.D.3 Safety System Status Monitoring Thatcher RES/DE/MEB NOTE 3(b) 7 06/30/95 NA I.D.4 Control Room Design Standard Thatcher RES/DRPS/RHFB NOTE 3(b) 7 06/30/95 NA I.D.5 Improved Control Room Instrisnentation Research I.D.5(1)

Operator-Process Cossaunication Thatcher RES/DF0/WBR NOTE 3(b) 7 06/30/95 NA I.D.5(2)

Plant Status and Post-Accident Monitoring Thatcher RES/DF0/W BR NOTE 3(a) 7 06/30/95 I.D.5(3)

On-Line Reactor Surveillance System Thatcher RES/DE/ME8 NOTE 3(b) 7 06/30/95 NA I.D.5(4)

Process Monitoring Instrumentation Thatche-RES/DF0/ICBR NOTE 3(b) 7 06/30/95 NA I.D.5(5)

Disturbance Analysis Systems Thatcher RES/DRPS/RHFB LI (NOTE 3)

F 06/30/95 NA I.D.6 Technology Transfer Conference Thatcher RES/DFC/HFBR LI (NOTE 3) 7 06/30/95 NA M

ANALYSIS AND DISSENINATION OF OPERATING EXPERIENCE I.E.1 office for Analysis and Evaluation of Operational Matthews AE00/PTB LI (NOTE 3) 2 06/30/95 NA Data y

zc I.E.2 Program Office Operational Data Evaluation Matthews NRR/DL/0RAB LI (NOTE 3)

CG/30/95 NA I.E.3 Operational Safety Data Analysts Matthews RES/DRA/RRBR LI (NOTE 3) 2 06/30/95 NA 7

c7 I.E.4 Coerdination of Licensee, Industry, and Regulatory Matthews AE00/PTB LI (NOTE 3) 2 06/30/95 NA Programs O

e I.E.5 Nuclear Plant Reitability Data System Matthews AE00/PTB LI (NOTE 3) 2 06/30/95 NA "w

1.E.6 Reporting Requirements P'.at thews AE00/PTB LI (NOTE 3) 2 06/30/95 NA U

!.E.7 Foreign Sources Matthews IP LI (NOTE 3) 2 06/30/95 NA O

O O

O O

O Table II (Continued)

Action Lead Office /

Safety Latest

$ Plan Itan/

Priority Division /

Priority Latest Issuance IFA g Issue No.

Title Engineer Branch Ranking Revision Date No.

?

E I.E.8 Human Error Rate Analysis Matthews RES/DF0/T8R LI (NOTE 3) 2 06/30/95 NA M

SIALITY ASSLRANCE I.F.I Expand QA List Pittman RES/DRA/AR618 NOTE 3(b) 2 06/30/89 NA I.F.2 Develop Nre Detailed QA Criteria I.F.2(1)

Assure the Independence of the Organization Performing Pittman DIE /DQASIP/QUAB LOW 2

06/30/89 NA the Checking Function I.F.2(2)

Include QA Personnel in Review and Approval of Plant Pittman DIE /DQASIP/QUA8 NOTE 3(a) 2 06/30/89 NA Procedures I.F.2(3)

Include QA Personnel in All Design, Construction, Pittman OIE/nQA$tr/QUA8 NOTE 3(a) 2 06/30/89 NA Installation, Testing, and Operation Activities I.F.Z'4)

Establish Criteria for Determining QA Requirements Pittman CIE/DQASIP/QUAS LOW 2

06/30/89 NA for Specific Classes of Equipment I.F.2(5)

Establish Qualification Requirements for QA and QC Pittman CIE/00ASIP/QUA8 LL'd 2

06/30/89 NA Personnel I.F.2(6)

Increase the Size of Licensees' QA Staff Pittman OIE/DQASIP/QUA8 NOTE 3(a) 2 06/30/89 NA

!.F.2(7)

Clarify that the QA Program Is a Condition of the Pittman OIE/DQASIP/QUA8 LOW 2

06/30/89 NA Construction Permit and Operating License I.F.2(8)

Compare NRC QA Requirements with Those of Other Pittman OIE/DQASIP/QUA8 LOW 2

06/30/89 NA Agencies U

I.F.2(9)

Clarify Organizational Reporting Levels for the QA Pittman OIE/DQASIP/QUA8 NOTE 3(a) 2 06/30/89 NA Organization I.F.2(10)

Clarify Requirements for Maintenance of "As-8uilt" Pittman O!E/DQASIP/QUA8 LOW 2

06/30/C9 NA Documentation I.F.2(11)

Define Role of QA in Design and Analysis Activities Ptttman DIE /DQASIP/QUAB LOW 2

06/30/89 NA y

PREOPERATIONAL AND LOW-POWER TESTING I.G.1 Training Requirements NRR/DHFS/PSR8 I

2 06/30/89 I.G.2 Scope of Test Program V'Molen NRR/DHFS/PSR8 NOTE 3(a) 2 06/30/89 NA y

SITING II.A.1 Siting Policy Reformulation V'Molen NRR/DE/SAE NOTE 3(b)

I 12/31/84 NA II.A.2 Site Evaluation of Existing Facilities V*Molen NRR/DE/SA8 V.A.1 1

12/31/84 NA M

CONSIDERATION OF DEGRADED OR MEtTED COPES IN II.6.1 Reactor Coolant System Vents NRR/DL I

3 12/31/91 F-10 11.8.2 Plant Shielding to Provide Access to Vital Areas and NRR/DL I

3 12/31/91 F-Il m

Protect Safety Equipment for Post-Accident Operation M

II.B.3 Post-Accident Sampling NRR/DL I

3 12/31/91 F-12 II.8.4 Training for Mitigating Core Demage 3 ERR /DL I

3 12/31/91 F-13 a

11.8.5 Research on Phenomena Associated with Core Degradation O

and Fuel Melting w

11.8.5(1)

Behavior of Severely Damaged Fuel V*Molen RES/DSR/AE8 LI (NOTE 5) 3 12/31/91 NA N

~

11.8.5(2)

Behavior of Core-Melt V'Molen RES/DSR/AE8 LI (NOTE 5) 3 12/31/91 NA

Table II fCentinued)

Action Lead Office /

Safety Latest om Plan Item /

Priority Olvision/

Priority Lstest Issuance IFA D

Issue No.

tit 1e Engineer Branch Ranking Revis1on Date No.

?.

m 11.8.5(3)

Effect of tfydrogen Burning and Explosions on V'Molen RES/DSR/AEB LI (NOTE 5) 3 12/31/91 NA Contairunent Structure 11.8.6 Risk Ra&ction for Operating Reactors at Sites with Pittman NRR/ DST /RRAB NOTE 3(a) 3 12/31/91 High Populat1on Dens 1 ties 11.8.7 Analysts of Hydrogen Control Matthews NRR/DSI/CS8 II.8.8 3

12/31/91 11.8.8 Rular:ktng Proceeding on Degraded Core Accidents V*Molen RES/DRA0/R8 Jet NOTE 3(a) 3 12/31/91 M

RELIABILITY ENGINEERING AND RISK ASSESSMENT II.C.1 Interim Reliability Evaluation Progree Pittman RES/DRAD/RR8 NOTE 2(o) 2 12/31/88 NA II.C.2 Continuation of Interim Reliability Evaluation Program Pittman NRR/ DST /RRA8 NOTE 3l5)

?

12/31/88 NA II.C.3 Systems Interaction Pittman NRR/ DST /GIB A-17 2

12/31/88 NA II.C.4 Reliability Engineering Pittman RES/DRPS/RHFB NOTE 3(b) 2 12/31/88 NA M

REACTOR COOLANT SYSTEM RELIEF AND SAFETY VAtVES II.D.1 Testing Requirements NRR/DL I

1 06/30/89 F-14 II.D.2 Research on Reitef and Safety Valve Test Requirements Riggs RES LOW 1

06/30/89 NA II.O.3 Relief and Safety Valve Position Indication NRR I

1 06/30/89 M

SYSTEM DESIGN g

4 II E.1 AuxtitarY Feedsater System II.E.1.1 Auxtttary Fee < hater System Evaluation NRR/DL I

1 12/31/86 F-15 II.E.1.2 Auxtitary Feedwater System Autenatic Initiation and NRR/DL I

1 12/31/86 F-16, F-17 Flow Indication

'!I.E.1.3 Update Standard Review Plan and Develop Regulatory Riggs RES/0RA/RRBR NOTE 3(a) 1 12/31/86 Guide 11.E 2 Emeroency Core Coolino System II.E.2.1 Reltance on ECCS Riggs NRR/DSI/RS8 II.K.3(17)

I 12/31/85 NA II.E.2.2 Research on Small Break LOCAs and Anomalous Transients Riggs RES/DAE/RSRB NOTE 3(b) 1 12/31/95 NA II.E.2.3 Uncertainties in Performance Predictions V'Malen NRR/DSI/RS8 LOW 1

12/31/85 NA 11 E 3 Decay Heat Removal II.E.3.1 Reitabiltty of Power Supplies for Natural Circulation NRR/DL I

1 06/30/91 II.E.3.2 Systems Reliability V'Malen NRR/ DST /GIB A-45 1

06/30/91 NA II.E.3.3 Coordinated Study of Shutdown Heat Removal Requirements V'Molen NRR/ DST /618 A-45 1

06/3C/91 NA II.E.3.4 Alternate Concepts Research Riggs RES/DAE/F8R8 NOTE 3(b) a 06/30/91 NA II.E.3.5 Regulatory Guide Riggs NRR/ DST /GIB A-45 1

06/30/91 NA (D

2 C

II E,4 Contatrument Desf on II.E.4.1 Dedicated Penetrations NRR/DL I

06/30/88 F-18 7

cn II.E.4.2 Isolation Dependability NRR/DL I

06/30/88 F-19 6

II.E.4.3 Integrity Check Nilstead RES/DRPS/RPSI NOTE 3(b) 06/30/88 NA O

e II.E.4.4 Purging II.E.4.4(1)

Issue Letter to Licensees Requesting Limited Purging Mtistead NRR/DSI/CS8 NOTE 3(a) 06/30/88 W

w 9

O O

N Table II (Continued)

Action Lead Office /

Safety Latest gm Plan Itan/

Priority Division /

Priority latest Issuance IFA g Issue No.

Yttle Engineer Branch Ranking Revision Date No.

?.

m II.E.4.4(2)

Issue Letter to Licensees Requesting Information on Mtistead NRR/DSI/ CSS NOTE 3(a) 06/30/88 Isolation Letter II.E.4.4(3)

Issue Letter to Licensees on valve Operablitty Mtistead NRR/DSI/CSB NOTE 3(a) 06/30/88 II.E.4.4(4)

Evaluate Purging and Venting During Normal Operation Milstead NRR/DSI/CSB NOTE 3(b) 06/30/88 NA II.E.4.4(5)

Issue Modified Purging and Venting Requirenent Mtistead NRR/DSI/CS8 NOTE 3(b) 06/30/88 NA II.E.5 Desten Sensttivity of 88W Reactors 11.E.5.1 Design Evaluation Thatcher NRR/DSI/R$8 NOTE 3(a) 1 12/31/84 II.E.5.2 88W Reactor Transtent Response Task Force Thatcher NRR/DL/0RAB NOTE 3(a) 1 12/31/84 II.E.6 In Situ Testino of Valves II.E.6.1 Test M e y Study Thatcher RES/DE/EIB NOTE 3(a) 1 06/30/89 M

INSTRtp H TATION AND CONTROLS II.F.1 Additional Accident Monitoring Instruentation NRR/DL I

2 06/30/89 F-20, F-21 F-22. F-23 F-24 F-25 II.F.2 Identification of and Recovery from Conditions NRR/DL I

2 06/30/89 F-26 Leading to Inadequate Core Cooling II.F.3 Instr aumts for Monitoring Accident Conditions V'Molen RES/DF0/ICBR NOTE 3(a) 2 06/30/89 II.F.4 Study of Control and Protective Action Design Thatcher NRR/DSI/ICSB DROP 2

06/30/89 NA Requirements II.F.5 Classification of Instrumentation Control, and Thatcher RES/DE LI (NOTE 3) 2 M/30/89 NA Electrical Equipment M

ELECTRICAL POWER 11.6.1 Power Suppites for Pressurtzer Reitef Valves, Block NRR I

12/31/94 NA Valves, and Level Irdicators M

TMI-2 CLEANUP A9D EXAMINATION II.H.1 Maintain Safety of TMI-2 and Ministre Envirorumental Matthews NRR/TMIPO NOTE 3(b) 2 06/30/95 NA Impact II.H.2 Obtain Technical Data on the Conditions Inside the Milstead RES/0RAA/AEB NOTE 3(b) 2 06/30/95 NA TMI-2 Containment Structure II.H.3 Evaluate and Feed Back Information Obtained from TMI M11 stead NRR/TMIPO II.H.2 2

06/30/95 NA II.H.4 Determine Impact of TMI on Socioeconomic and Real Mtistrad RES/DHSWM/SE:ER LI (NOTE 3) 2 06/30/95 NA Property Values

'x3

$M GENERAL : MPtICAT ONS 7 TMI FOR DESIGL AND g

CONSTRUC Ian ACT VITl [5 g

o O

8 II.J l Vendor Inspection Proaran 3

O II.J.1.1 Estabitsh a Priority System for Conducting Vendor Riant OIE/DQASIP LI (NOTE 3) 11/30/83 NA N

W Inspections

~

II.J.I.2 Modify Existing vendor Inspection Program Riant DIE /DQASIP LI (NOTE 3) 11/30/83 NA

Table II (Continued)

Action o

Lead Office /

Safety Latest m Plan Item /

Priority Division /

Priority latest Issuance MPA D 1: sue No.

Title Engineer Branch Ranking Revision Date No.

O W

cn II.J.l.3 increase Regulatory Control Over Present Non-Licensees Riant OIE/DQASIP LI (NOTE 3) 11/30/83 NA II.J.l.4 Assign Resident Inspectors to Reactor Vendors and Riant DIE /DQASIP LI (NOTE 3) 11/30/83 NA Architect-Engineers II.J,2 Construction Innection Proorse

!!.J.2.1 Reortent Construction Inspection Program Riant DIE /DQASIP LI (NOTE 3) 11/30/83 NA II.J.2.2 Increase Eghasis on Independent Measurement in Riant DIE /DQASIP LI (NOTE 3) 11/30/83 NA Construction Inspection Program II.J.2.3 Assign Restdent Inspectors to All Construction Sites Riant DIE /DQASIP L1 (NOTE 3) 11/30/83 NA II.J.3 Manaoement for Deston and Construction II.J.3.1 Oiganizatton and Staffing to Oversee Design and 5tttman NRP/DHFS/LQB I.B.I.I 11/30/83 NA Construction II.J.3.2 Issue Regulatory Guide Pittman NRR/DHFS/LQB I.B.I.1 11/30/84 NA II.J.4 Revise Deficiency Reportino Reautrements II.J.4.1 Revise Deficiency Reporting Requirements Riani AE00/DSP/ROAB NOTE 3(a) 2 06/30/95 NA M

MEASLRES TO MITIGATE SMAlt-BR EAK TOSS-OF-COOLANT AllDtuib AND 1055-OF-FttDtfATTR ACCIDtRib II.K.!

IE Bulletins II.K.l(l)

Review TMI-2 PNs and Detailed Chronology of the Emrit NRR NOTE 3(a) 12/31/84 TMI-2 Accident II.K.l(2)

Review Transients Statlar to TNI-2 That Have Eartt NRR NOTE 3(a) 12/31/84 Occurred at Other Factittles and NRC Evaluation of Davis-Besse twent li.K 1(3)

Review Operating Procedures for Recognizing, Emrit NRR NOTE 3(a) 12/31/84 Preventing, and Mitigating Void Formation in Transtents and Accidents II.K.l(4)

Review Operating Procedures and Training Eerit NRR NOTE 3(a) 12/31/84 Instructions II.K.l(S)

Safety-Related Valve Position Description Eerit NRR NOTE 3(a) 12/31/84 II.K.l(6)

Review Contatnnent Isolation Initiation Design Eartt NitR NOTE 3(a) 12/31/84 and Procedures II.K.l(7)

Implement Positive Position Controls on Valves Eerit NRR NOTE 3(a) 12/31/84 That Could Co g ro=tse or Defeat AFW Flow II.K.l(8)

Iglement Procedures That Assure Two Independent Emrit NRR NOTE 3(a) 12/31/84 100% AFV Flow Paths II.K.l(9)

Review Procedures to Assure That Radioactive Emrit NRR NOTE 3(a) 12/31/84 Liquids and Gases Are Not Transferred out of y

,E Containment Inadvertently Q

II.K.l(10)

Review and Modify Procedures for Removing Safety-Enrit CR NOTE 3(a) 12/31/84 7

m Related Systens from Service g

II.K.l(II)

Nake All Operating and Maintenance Personnel Emrit NRR NOTE 3(a) 12/31/84

-3 e

Aware of the Seriousness and Consequences of the y

Erroneous Actions Leading up to, and in Early U

Phases of, the TNi-2 Accident O

O O

m.

\\

(

V s

Table II (Continued)

Action Lead Offtce/

Safety Latest

$ Plan Item /

Priority Olvision/

Priority Latest Issuance MPA g Issue No.

Title Engineer Branch Ranking Revision Date No.

?

$ II.K.l(12)

One Hour Notification Requirement ard Continuous Eerit NRR NOTE 3(a) 12/31/84 rm ications Channels II.K.1(13)

Propose Technical Spectftcailan Changes Reflecting Enrtt NRR NOTE 3(a) 12/31/84 Implementation of All Bulletin Items II.K.1(14)

Review Operating Modes and Prncedures to Deal with Eerit NRR NOTE 3(a) 12/31/64 Significant Amounts of Hydrogen II.K.1(15)

For Facilities with ison-Automatic AFV Initiation.

Eurit NRR NOTE 3(a) 12/31/84 Provide Dedicated Operator in Continuous Comunsntcation with CR to Operate AFV II.K.1(16) laplement Procedures That Identify PRI PORY "Open" Eartt NRR NOTE 3(a) 12/31/84 Indications and That Direct Operator to Close Manually at " Reset" Setpoint II.K.1(17)

Trip PZR Level Sistable so That PZR Law Pressure Eartt NRR NOTE 3(a) 12/31/84 if111 Initiate Safety Injection II.K.l(18)

Develop Procedures and Train Operators on Methods Eartt istR NOTE 3(a) 12/31/84 of Estabitshing and Matataining Natural Circulation II.K.l(19)

Describe Design and Procedure Modifications to Enrtt NRR NOTE 3(a) 12/31/84 Reduce L1kelIhood of Automatic PZR PORV Actuation in Transtants II.K.1(20)

Provide Proce & res and Training to Operators for Enrtt NRR NOTE 3(a) 12/31/84 Prompt Manual Reactor Trip for LOFW, TT, MSIV d

Closure, LOOP, LOSG Level, and L0 PZR Level II.K.1(21)

Provide Automatic Safety-Grade Anticipatory Reactor Enrit NP2 NOTE 3(a) 12/31/84 Trip for LOFV, TT, or Significant Decrease in SG Level II.K.1(22)

Describe Automatic and Manual Actions for Proper Eartt NRR NOTE 3(a) 12/31/84 Functioning of Auxtilary Heat Removal Systems When FV System Not Operable

!!.K.1(23)

Describe Uses and Types of RV Level Indication for Eartt NRR NOTE 3(a) 12/31/84 Automatic and Manual Initiation Safety Systems II.K.1(24)

Perforu LOCA Analyses for a Range of Small-Break Eerit NRR NOTE 3(a) 12/31/84 Stres and a Range of Time Lapses Between Reactor Trip and RCP Trip II.K.1(25)

Develop Operator Action Guidelines Ecrit NRR NOTE 3(a) 12/31/64 II.K.1(26)

Revise Emergency Procedures and Train R0s and SR0s Eerit NRR NOTE 3(a) 12/31/64 II.K.l(27)

Provide Analyses and Develop Guidelines and Eartt NRR NOTE 3(a) 12/31/84 Procedures for Inadequate Core Cooling Conditions II.K.l(28)

Provide Design That Will Assure Automatic RCP Trip Enrit NRR NOTE 3(a) 12/31/84 for All Circumstances Where Required II.K.2 Commission Orders on 88W Plants II.K.2(1)

Upgrade Timeliness and Reitablitty of AFV System Eurit NRR/DSI NOTE 3(a) 12/31/84 c-II.K.2(2)

Procedures and Training to Initiate and Control Enrit NRR NOTE 3(a) 12/31/84 y

n AFV Independent of Integrated Control Systen Q II.K.2(3)

Hard-Wired Control-Grade Anticipatory Reactor Trips Enrtt Intr /DSI NOTE 3(a) 12/31/84 p

g II.K.2(4)

Snell-Break LOCA Analysis, Procedures and Operator Enrit IstR/DeFS/0LB NOTE 3(a) 12/31/84 Q

e Training y II.K.2(S)

Complete TMI-2 Simulator Training for All Operators Emrtt NRR NOTE 3(a) 12/31/84 II.K.2(6)

Reevaluate Analysis for Dual-Level Setpoint Control Enrtt NRR/DSI NOTE 3(a) 12/31/84

Table Il (Continued 1 Action Lead Office /

Safety Latest

$ Plan Item /

Priority Olvision/

Priority Latest Issuance WA D issue No.

Title Engineer Branch Ranking Revision Date No.

o cn II.K.2(7)

Reevaluate Transtent of Septem6er 24, 1977 Eartt NRR/DSI NOTE 3(a) 12/31/84 II.K.2(8)

Continued upgrading of AFW Systen Emrit NRR II.E.1.1, 12/31/84 NA II.E.1.2 II.K.2(9)

Analysis and Upgrading of Integrated Control System Enrtt NRR I

12/31/84 F-17 II.K.2(10)

Hard-Wired Safety-6rade Anticipatory Reactor Trips Emrit NRR I

12/31/84 F-28 II.K.2(II)

Operator Training and Drt11tng Eerit NRR I

12/31/84 F-29 II.K.2(12)

Transient Analysis and Procedures for Management Enrit NRR I.C.1(3) 12/31/84 NA of Small Breaks II.K.2(13)

Theriaal-Mechanical Report on Effect of HPI on vessel Enrit NRR I

12/31/84 F-30 Integrity for Small-Break LOCA With No AFV II.K.2(14)

Damonstrate That Predicted Lift Frequency of PORVs Enrit NRR I

12/31/84 F-31 and SVs Is Acceptable II.K.2(15)

Analysis of Effects of Slug Flow on once-Tnrough Emrit NRR I

12/31/84 Steam Generator Tubes After Primary System Voiding II.K.2(16)

Ispact of RCP Seal Damage following Small-Break Enrit NRR I

12/31/84 F-32 LOCA With Loss of Offsite Power II.K.2(17)

Analysts of Potential Volding in RCS During Eerit NRR I

12/31/64 F-33 Anticipated Transients II.K.2(18)

Analysts of Loss of Feedwater and Other Anticipated Eerit NRR I.C.l(3) 12/31/84 NA Transients II.K.2(19)

Benchmark Analysts of Sequential AFV Flow to once-Eartt NRR I

12/31/84 F-34 Through Steam Generator m

ll.K.2(20)

Analysts of Steam Response to Small-Break LOCA Enrit NRR I

12/31/84 F-35 That Causes System Pressure to Exceed PORV Setpoint II.K.2(21)

LOFT L3-1 Predtettons Emrit NRR/DSI NOTE 3(a) 12/31/84 II.K.3 Final Reconsmendations of Bulletins and Orders Task Force II.K.3(1)

Install Automatic PORV lsolation System and Perform Emrit NRR I

12/31/84 F-36 Operational Test II.K.3(2)

Report on Overall Safety Effect of PORV Isolation Eerit NRR I

12/31/84 F-37 System II.K.3(3)

Report Safety and Relief Valve Failures Prongtly Emrit NRR I

12/31/84 F-38 and Challenges Annually II.K.3(4)

Review and Upgrade Reliability and Redundancy of Emrit NRR

!!.C.1, 12/31/84 NA Non-Safety Equipment for Small-Break LOCA Nitigation II.C.2, II.C.3 II.K 3(S)

Automatic Trio of Reactor Coolant Pteps Emrit NRR I

12/31/84 F-39, G-01 II.K.3(6)

Instrumentation to Verify Natural Circulation Emrit NRR/DSI l.C.l(3),

12/31/84 NA II.F.2, II.F.3 II.K.3(7)

Evaluation of PORV Opening Probablitty During Eart t' NRR I

12/31/84 c-Overpressure Transtent

U II.K.3i8)

Further Staff Consideration of Need for Diverse Emrit NRR/ DST /GIB II.C.1, 12/31/84 NA 7

Decay Heat Removal Method Independent of SGs II.E 3.3 8

!!.K.3(9)

Proportional Integral Derivative Controller Eerit NRR I

12/31/84 F-40 Modification W

w W

O O

O

- Table Il (Continued)

Action Lead Office /

Safety Latest g Plan item /

Priority Division /

Priority latest Issuance IFA Issue IIo Title Engineer Branch Ranking Revision Date No.

w IF.K.3(10)

Antletpatory Trip Modification Proposed by Some Eartt NRR I

12/31/84 F-41 Licensees to Confine Renge of Use to High Power Levels II.K.3(11)

Control Use of PORV Supplied by Control Components.

Eartt NRR I

12/31/84 Inc. thttl Further Review Complete II.K.3(12)

Confim Existence of Anticipatory Trip Upon Turbine Eartt IIRR I

12/31/84 F-42 Trip II.K.3(13)

Separetton of IfCI and RCIC System Initiation tevels Eerit NRR I

12/31/84 F-43 II.K.3(14)

Isolation of Isolation Condensers on High Radiation Eartt flRR I

12/31/84 F-44 II.K.3(15)

Modify Break Detection Logic to Prevent Spurious Enrit NRR I

12/31/84 F-45 Isolation of 19C1 and RCIC Systems II.K.3(16)

Reduction of Challenges and Failures of Relief Eartt NRR I

12/31/84 F-46 Valves - Feasiblitty Study and System leodification II.K.3(IT)

Report on Outage of ECC Systems - Licensee Report Eartt IIRR I

12/31/84 F-47 and Technical Specification Changes II.K.3(18)

Modification of ADS Logic - Feasiblitty Study and Enrit NRR 1

12/31/84 F-48 Hodification for Increased Diversity for Some Event Sequences II.K.3(19)

Interlock on Rectreulation Pimp Loops Eartt NRR I

12/31/84 F-49 II.K.3(20)

Loss of Service Water for Big Rock Point Enrit NRR I

12/31/84 II.K.3(21)

Restart of Core Spray and LPCI Systems on Low Enrit NRR I

12/31/84 F-50 Level - Design and ModtfIcation II.K.3(22)

Automatic Switchover of RCIC System Suction -

Eerit NRR I

12/31/84 F-51 Vertfy Procedures and Modify Design II.K.3(23)

Central Water Levei Recording Enrit NRR I.D.2 12/31/84 NA III.A.I.2(1).

III.A.3.4 II.K.3(24)

Confim Adequacy of Space Cooling for HPCI and Eerft NRR I

12/31/84 F-52 RCIC Systems II.K.3(25)

Effect of Loss of AC Power on Pisup Seals Enrit NRR

~

I 12/31/84 F-53 II.K.3(26)

Study Effect on Rigt Reliablitty of Its Use for Enrit NRR/DSI II.E.2.1 12/31/84 NA Fuel Pool Cooling II.K.3(27)

Provide Consnon Reference Level for Vessel Level Emrit NRR I

12/31/84 F-54 Instrismentation II.K.3(28)

Study and Verify Qualification of Accumulators Emrit NRR I

12/31/84 F-55 on A05 Valves II.K.3(29)

Study to Demonstrate Performance of Isolation Eurit NRR I

12/31/84 F-56 Condensers with flon-Condensibles II.K.3(30)

Revised Small-Break LOCA Methods to Show Compitance Enrtt NRR I

12/31/84 F-57 with 10 CFR 50. Appendix K

x3 g

II.K.3(31)

Plant-Spectf tc Calculations to Show Compilance with Enrtt NRR I

12/31/84 F-58 m

10 CFR 50.46 g

II.K.3(32)

Provide Experimental Verification of Two-Phase Emrit NRR/DSI II.E.2.2 12/31/84 NA '

e Natural Circulation Models O

g II.K.3(33)

Evaluate Elimination of PORV Function Emrit NRR II.C.1 12/31/84 NA

~3 w

II.K.3(34)

Relap-4 Model Development Emrit NRR/DSI II.E.2.2 12/31/84 MA N

~

Table 11 fContinuedi Actton iead Office /

Safety Latest om Plan Item /

Priority Division /

Priority latest Issuance IFA D Issue No.

Title Engineer Branch Ranking Revision Date No.

o N

40

• II.K.3(35)

Evaluation of Effects of Core Flood Tank Injection Enrit NRR 1.C.1(3J 12/31/84 NA on Sas11-Break LOCAs I!.4.1(36)

Additional Staff Audit Calculations of 84W Sunil-Eartt NRR I.C.1(3) 12/31/84 NA Break LOCA Analyses II.K.3()T)

Analysts of 88W Response to Isolated Small-Break Enrtt NRR I.C.1(3) 12/31/84 NA LOCA II.K.3(38)

Analysis of Plant Response to a Small-Break LOCA in Enrit NRR I.C.1(3) 12/31/84 NA the Pressurizer Spray Line II.K.3(39)

Evaluation of Effects of Water Slugs in Piping Eartt NRR I.C.1(3) 12/31/84 NA Caused by 191 and CFT Flows II.K.3(40)

Evaluation of RCP 5eal Damage and Leakage During Emrtt NRR II.K.2(16) 12/31/84 NA a Smell-Break LOCA II.K.3(41)

Submit Predictions for LOFT Test L3-6 with RCPs Eartt istR I.C.1(3) 12/31/84 NA Running II.K.3(42)

Submit Requested Information on the Effects of Emrtt NRR I.C.1(3) 12/31/84 NA non-Condensible 6ases II.K.3(43)

Evaluation of feechanical Effects of Slug Flow on Enrtt NRR II.K.2(15) 12/31/84 NA Steam Generator Tubes II.K.3(44)

Evaluation of Anticipated Transtents with single Eartt letR I

12/31/84 F-59 Failure to Verify No Significant Fuel Failure II.K.3(45)

Evaluate Depressurization with Other Than Full A05 Emrit NRR I

12/31/84 F-60

,o II.K.3(46)

Response to List of Concerns from ACRS Consultant Eartt NRR I

12/31/84 F-61 II.K.3(47)

Test Progrme for Samil-Break LOCA lbdel Vertf tcation Enrtt NRR I.C.1(3),

12/31/84 NA Pretest Prediction. Test Program, and Model II.E.2.2 Vertitcation II.K.3(48)

Assess Change in Safety Reliability as a Result of Eartt NRR II.C.1, 12/31/84 NA laplementing 840TF Reconsiendations II.C.2 II.K.3(49)

Review of Procedures (NRC)

Enrtt NRR/DHF5/PSR8 I.C.8 12/31/84 NA I.C.9 II.K.3(50)

Review of Procedures (N555 vendors)

Eartt NRR/Def5/PSRB I.C.T.

12/31/84 NA I.C.9 II.K.3(51)

Symptom-Based Emergency Procedures Enrit NRR/DHF5/PSR8 I.C.9 12/31/84 NA II.K.3(52)

Operator Amareness of Revised Emergency Procedures Enrit fatR I.8.1.1, 12/31/84 NA I.C.2 I.C.5 II.K.3(53)

Tuo Operators in Control Room Enrtt NRR I.A.1.3 12/31/84 NA II.K.3(54)

Simulator Upgrade for Small-Break LOCAs Eerit NRR I.A.4.l(2) 12/31/84 NA II.K.3(55)

Operator Monitoring of Control Board Enrit NRR I.C.1(3),

12/31/84 NA I.D.2 I.D.3 II.K.3(56)

Simulator Training Requirements Curit NRR/ DES /0LB I.A.2.6(3),

12/31/84 NA y

2c:

I.A.3.1 ac y

II.K.3(57)

Identify Water Sources Prior to Manual Activatton Eerit NRR I

12/31/84 F-62 7

of ADS p

g 3

o O

O O

m m _

f J

(

\\

Table II (Continued)

Action Lead Office /

Safety Latest

$ Plan Item /

Priority Division /

Priority Latest Issuance frA g Issue No.

Title Engineer Branch Raniting Revision Date No.

$ Q EERGEllCY PREPAREDNESS AND ItADIATION EFFECTS III.A.1 leprove Licensee Emereency Prenaro6ess - Short-Tens III.A.1.1 Upgrade Emergency Prepare & ess Ill.A.I.1(1)

Implement Action Plan Requirements for Promptly OIE/DEPER/EP8 I

2 06/30/91 Ieproving Licensee Emergency Prepare & ess III.A.I.1(2)

Perfone an Integrated Assessment of the Implementation DIE /MPER/EP8 NOTE 3(b) 2 06/30/91 NA III.A.I.2 Upgrade Licensee Emergency Support Factitties 2

06/30/91 III.A.I.2(1)

Technical Support Center DIE /DEPER/EPS I

2 06/30/91 F-63 III.A.1.2(2)

On-Site Operational Support Center CIE/DEPER/EPB I

2 06/30/91 F-64 III.A.I.2(3)

Near-$tte Emergency Operations Fact 11ty OIE/DEPER/EP8 I

2 06/30/91 F-65 III.A.I.3 Maintain Supplies of Thyroid-Blociting Agent 2

06/30/91 111.A.1.3(1)

Wrters Riggs CIE/DEPER/EPS NOTE 3(b) 2 06/30/91 NA III.A.I.3(2)

Pubile Riggs DIE /DEPER/EP8 IIOTE 3(b) 2 06/30/91 NA III.A.2 Imerovina Licensee Emeroency Precare &ess - Lono-Tens III.A.2.1 Amend 10 CFR 50 and 10 CFR 50 Appendix E III.A.2.1(1)

Pubitsh Proposed Amendments to the Rules RES NOTE 3(a) 12/31/94 NA

[

llt.A.2.1(2)

Conduct Pubite Regional Nmetings RES IIGTE 3(b) 12/31/94 NA f

III.A.2.1(3)

Prepare Final Commission Paper Recommending Adoption RES IIOTE 3(b) 12/31/94 IlA of Rules i

i III.A.2.1(4)

Revise Inspection Program to Cover Upgraded OIE I

F-67 Requiraments III.A.2.2 Development of Guidance and Criteria NRR/DL I

F-68 III.A.3 Iserovina NRC Emeraency Preparedness III.A.3.1 estC Role in Responding to fluclear Emergencies III.A.3.1(1)

Define NRC Role in Emergency situations Riggs CIE/DEPER/IRD6 IIOTE 3(b) 1 06/30/85 NA Ill.A.3.1(2)

Revise and Us, grade Plans and Procedures for the NRC Riggs O!E/ DEFER /IRD8 h0TE 3(b) 1 06/30/85 NA Emergency Operations Center III.A.3.l(3)

Revise Manual Chapter 0502, Other Agency Procedures, Riggs OIE/DEPER/IRDB NOTE 3(b) 1 06/30/85 NA and NURES-0610

!!I.A.3.l(4)

Prepare Commission Paper Riggs CIE/DEPER/IRDB NOTE 3(b)

I e6/30/85 NA III.A 3.l(5)

Revise Implementing Procedures and Instructions for Riggs OIE/DEPER/IRDB NOTE 3(b) 1 06/30/85 NA Regional Offices Ill.A.3.2 Improve Operations Centers Riggs CIE/DEPER/IRD6 MOTE 3(b) 1 06/30/85 MA III.A.3.3 Communications III.A.3.3(1)

Install Direct Dedicated Telephone Lines Pittman OIE/DEPER/IRDe NOTE 3(a) 1 06/30/85 NA III.A.3.3(2)

Obtain Dedicated, Short-Range Radio Communication Pittman CIE/DEPER/IRD8 IIOTE 3(a) 1 06/30/85 MA Systems III.A.3.4 Nuclear Data Link Thatcher die /DEPER/IRDB IIOTE 3(b) 1 06/30/85

U III.A.3.5 Training. Drills, and Tests Pittman OIE/DEPER/IRD8 NOTE 3(b) 1 06/30/85 NA l

o III.A.3.6 Interaction of IstC and Other Agencies

-^

Q III.A.3.6(1)

Internettonal Pittman OIE/DEPER/EPLB IIOTE 3(b) 1 06/30/85 NA a

III.A.3.6(2)

Federal Pittman OIE/DEPER/EPLB NOTE 3(b) 1 06/30/85 NA C)

III.A.3.6(3)

State and Local Pittman CIE/DEPER/EPLB IIOTE 3(b) 1 06/30/85 NA w

N W

I Table II (Continuadi Action Lead Office /

Safety latest Plan Item /

Priority Division /

Priority latest Issuance MPA 0,

N Issue No.

Title Engineer Branch Ranking Revision Date No.

N$ Q EhERGENCY PREPAREDNESS OF STATE AND t0 CAL GOVERNMENTS III.B.1 Transfer of Responsthtllties to FEMA Milstead DIE /DEPER/IRDB NOTE 3(b) 11/30/83 NA 111.8.2 Implementation of lutt and FEMA Responsibilities 111.8.2(1)

The Licensing Process Nilstead 01E/DEPER/IRDS NOTE 3(b) 11/30/83 NA 111.8.2(2)

Federal Guidance Milstead DIE /DEPER/lRD8 NOTE 3(b) 11/30/83 NA Q

PUBLIC INFORMATION III.C.1 Have Information Avallable for the News Media and the PublIc III.C.ltt)

Review Pditely Available Doctaments Pittman PA LI (NOTE 3) 11/30/83 siA III.C.I(2)

Recommmend Pubitcation of Additional Information Pittman PA LI (NOTE 3) 11/30/83 NA III.C.1(3)

Program of Seminars for News Media Personnel Pittman PA LI (NOTE 3) 11/30/83 NA III.C.2 Develop Policy and Provide Training for Interfacing With the News Media Ill.C.2(1)

Develop Policy and Procedures for Dealing With Briefing Pittman PA LI (NOTE 3) 11/30/83 NA Requests III.C.2(2)

Provide Training for Members of the Technical Staff Pittman PA LI (NOTE 3) 11/30/83 NA Q

RADI ATION PROTECTION III.D 1 Radiation Source Control III.D.l.1 Primary Conlant Sources Outside the Containment Structure 111.D.1.l(1)

Review Information Submitted by Licensees Pertaining NRR I

1 12/31/&3 to Redacing Leakage from Operating Systems III.D.I.l(2)

Review Informtion on Provisions for Leak Detection Enrit RES/DRA/ARGIB DROP 1

12/31/88 111.D.1.1(3)

Develop Proposed System Acceptance Crtteria Enrit RES/DRA/ARGIB DROP 1

12/31/88 III.D.I.2 Radioactive Gas Management Emrit NRR/DSI/ MET 8 DROP 1

12/31/88 NA 111.D.1.3 Ventilation System and Radioiodine Adsorber Criteria III.D.I.3(1)

Decide Whether Licensees Should Perform Studies and Emrit NRR/DSI/ MET 8 DROP 1

12/31/88 NA Make Modifications III.D.I.3(2)

Review and Revise SRP Eerit NRR/DSI/METB DROP 1

12/31/88 NA 111.D.1.3(3)

Require Licensees to Upgrade Filtration Systems Emrit NRR/DSI/ MET 8 DROP 1

12/31/88 NA III.D.I.3(4)

Sponsor Studies to Evaluate Charcoal Adsorber Eerit NRR/DSI/METB NOTE 3(b) 1 12/31/88 NA 111.0 l.4 Radwaste System Design Features to Aid in Accident Eerit NRR/DSI/ MET 8 DROP 1

12/31/88 NA Recovery and Decontamination III D 2 Public Radiation Protection Isorovement III.D.2.1 Radiological Monitoring of Effluents

D 2._

lll.D.2.l(1)

Evaluate the Feasibility and Perform a Value-Impact Eerit NRR/DSI/METB LOW 2

12/31/85 NA c

D Analysis of Modifying Ef fluent-Monitoring Design Criteria s

III.D.2.l(2)

Study the Feasibility of Requiring the Development Eerit NRR/DSI/ MET 8 LOW 2

12/31/85 NA O

8 of Effective Means for Monitoring and Sampling Noble N

w Gases and Radiciodine Released to the Atmosphere

~

III.D.2.l(3)

Revtse Regulatory Guides Eartt NRR/DSI/ MET 8 LOW 2

12/31/85 NA W

O O

O

\\

Table II fContinued)

Action Lead Office /

Safety Latest

$ Plan Item /

Priority Olvision/

Priority Latest Issuance IFA g Issue No.

Title Engineer Branch Ranking Revision Date No.

?$

III.D.2.2 Radiciodine. Carbon-14. and Tritius Pathuey Dose Analysis III.D.2.2(1)

Perform Study of Radiciodine. Carbon-14, and Trittian Eartt NRR/DSI/RA8 NOTE 3(b) 2 12/31/85 NA Behavior III.D.2.2(2)

Evaluate Data Collected at Quad Cities Enrtt NRR/DSI/RA8 III.D.2.5 2

12/31/85 IIA III.D.2.2(3)

Determine the Distributton of the Chemical Species of Enrit lutR/DSI/RA8 III.D.2.5 2

12/31/85 NA Radiotodine in Atr-Water-Steam Ittrtures III.D.2.2(4)

Revise SRP and Regulatory Guides Enrtt NRR/DSI/RA8 III.D.2.5 2

12/31/05 NA III.D.2.3 Liquid Pathuey Radiological Control III.D.2.3(1)

Develop Procedures to Discriminate Between Emrtt lutR/DE/EHE8 NOTE 3(b) 2 12/31/85 NA Sites / Plants III.D.2.3(2)

Discriminate Between $ttes and Plants That Require Enrit NRR/DE/EE8 IIOTE 3(b) 2 12/31/85 NA Consideration of Liquid Pattesay Interdiction Techniques III.D.2.3(3)

Estabilsh Feasible Method of Pathway Interdiction Enrtt NRR/DE/EHE8 NOTE 3(b) 2 12/31/85 NA III.D.2.3(4)

Prepare a Sammunry Assessment Eartt NRR/DE/EHE8 NOTE 3(b) 2 12/31/85 NA III.D.2.4 Offstte Dose Measurements III.D.2.4(1)

Study Feastbtitty of Envirorumental Monitors V*Molen IIRR/DSI/RA8 NOTE 3(b) 2 12/31/85 NA III.D.2.4(2)

Place 50 TLDs Around Each Site V*Molen OIE/DRP/0RPS LI (NOTE 3) 2 12/31/85 NA III.D.2.5 Offsite Dose Calculation Manual V *Molen NRR/DSI/RA8 NOTE 3(b) 2 12/31/85 NA III.D.2.6 Independent Radiological Measurements V *Molen OIE/DRP/ORPS LI (NOTE 3) 2 12/31/85 NA O

III.D.3 Worker Radiation Protection Improvement III.D.3.1 Radiation Prot 2 tton Plans V *Molen NRR/DSI/RA8 NOTE 3(b) 3 12/31/87 MA III.D.3.2 Health Physics Improvements III.D.3.2(1)

Amend 10 CFR 20 V*Molen RES/DF0/ORPBR LI (NOTE 3) 3 12/31/87 NA III.D.3.2(2)

Issue a Regulatory Guide V*Molen RES/DF0/ORPOR LI (NOTE 3) 3 12/31/87 NA III.D.3.2(3)

Develop Standard Perfonmence Criterta V *Molen RES/DF0/ORP8R L1 (NOTE 3) 3 12/31/87 NA III.D.3.2(4)

Develop Method for Testing and Certifying Air-Purifying V *Molen RES/DF0/0RP8R LI (NOTC 3) 3 12/31/87 NA Respirators III.D.3.3 In-plant Radiation Monitoring III.D.3.3(1)

Issue Letter Requiring Improved Radiation Sampling NRR/DL I

2 F-69 Instrumentation III.D.3.3(2)

Set Criteria Requiring Licensees to Evaluate IIeed for sNtt IIOTE 3(a) 2 12/31/86 NA Additional Survey Equipement III.D.3.3(3)

Issue a Rule Change Providing Accestable Methods for RES NOTE 3(a) 2 12/31/86 NA Calibration of Radiation-Monitoring Instruments III.D.3.3(4)

Issue a Regulatory Guide RES IIOTE 3(a) 2 12/31/86 NA III.D.3.4 Control Room Habitability NRR/DL I

F-10 III.D.3.5 Radiation Worker Exposure III.D.3.5(1)

Develop Format for Data To se Collected by Uttittles V*Molen RES/DF0/0RP8R LI (NOTE 3) 2 12/31/86 NA Regarding Total Radiation Exposure to Workers g

c III.D.3.5(2)

Investigative Methods of obtaining Employee Health V*Iblen RES/DF0/ORPOR LI (NOTE 3) 2 12/31/86 MA M

Data by Nonlegislative Means E

111.D.3.5(3)

Revise 10 CFR 20 V *Molen RES/DF0/ORPOR LI (NOTE 3) 2 12/31/86 NA 1

o e

M STRENGTEN ENFORCEENT PROCESS m

~

IV.A.1 Seek Legislative Authority Eerit GC LI (NOTE 3) 11/30/83 NA

Table II fContinued)

Action Lead Office /

%fety Latest

$ Plan Itesn/

Priority Olvision/

Priority latest Issuance MPA g Issue No.

Title Engineer Branch Ranking Revision Date No.

k-IV.A.2 Revise Enforcement Policy Eerit DIE /ES LI (NOTE 3) 11/30/83 NA g

ISSUANCE OF INSTRUCTIONS AND INFORMATION TO LICENSEES IV.8.1 Revise Practices for Issuance of Instructions and Enrit O!E/DEPER L1 (NOTE 3) 11/30/83 NA information to Licensees M

EXTEN) L ESSONS LEARNED T0 tICENSED ACTIVITIES OTHER THAN Wlt REACHES IV.C.1 Extend Lessons Learned fran TMI to Other NRC Programs Enrit NMSS/WM NOTE 3(b) 11/30/83 NA g

NRC STAFF TRAINING IV.D.1 NRC Staff Training Eerit ADM/ MOTS LI (tsTE 3) 11/30/83 NA g

SAFETY DECISION-MAKING IV.E.1 Expand Research on Quantification of Safety Colmar RES/DRA/RABR LI (NOTE 3) 2 12/31/86 NA Decis1on-Making IV.E.2 Plan for Early Resolution of Safety Issues Enrit NRR/ DST /SPEB LI (NOTE 3) 2 12/31/86 NA IV.E.3 Plan for Resolving issues at the CP Stage Colmar RES/DRA/RABR LI (NOTE 5) 2 12/31/86 NA IV.E.4 Resolve Generic Issess by Rulemaking Colmar RES/DRA/RABR LI (NOTE 3) 2 12/31/86 NA IV.E.5 Assess Currently Operating Reactors Matthews NOR/DL/SEP8 NOTE 3(b) 2 12/31/86 NA M

FINANCTAL DISINCENTIVES TO SAFETY IV.F.1 Increased OIE Scrutiny of the Power-Ascension Test Thatcher DIE /DQASIP NOTE 3(b)

I 12/31/86 NA Program IV.F.2 Evaluate the Impacts of Financial Distncentives to Matthews SP NOTE 3(b) 1 12/31/86 NA the Safety of Nuclear Pcwr Plants IV G IMPROVE SAFETY RULEMAKING PROCEDURES IV.G.1 Develop a Public Agenda for Rulemaking Enn i t ADM/RPS LI (NOTE 3) 1 12/31/86 NA IV.G.2 Periodic and Systematic Reevaluation of Existing Rules Milstead RES/DRA/RABR LI (NOTE 3) 1 12/31/86 NA IV.G.3 Improve Rulemaking Procedures Milstead RES/DRA/RABR LI (NOTE 3)

I 12/31/86 NA IV.G.4 Study Alternatives for leproved Rulemaking Process Milstead RES/DRA/RABR L1 (NOTE 3) 1 12/31/86 NA g

NRC PARTICIPATION IN THE RADIATION POLICY COUNCIL

=

IV.H.1 NRC Participation in the Radiation Policy Council Sege RES/DHSWM/HEBR LI (NOTE 3) 11/30/83 NA k

o M

DEVELOPMENT OF SAFETY P0t!CY

."o V.A.)

Develop NRC Policy Statement on Safety Eartt GC LI (NOTE 3) 12/31/86 NA N

O O

O

)

Table II (Continued 1 Action Lead Office /

Safety Latest Plan Item /

Priority Otvision/

Priority Latest Issuance IFA g

Issue No.

Title Engineer Branch Ranking Revision Date No.

o

<n M

POSSIBLE ELININATION OF NORSAFETY RESPONSIBILITIES V.B.1 Study and Reconuend, as Appropriate. Elimination of Eartt GC LI (NOTE 3) 12/31/86 NA Nonsafety Responsib111ttes M

ADVISORY ColeIITTEIS V.C.1 Strengthen the Role of Advisory Committee on Reactor Eerit GC L1 (IIOTE 3) 12/31/86 NA Safeguards V.C.2 Study Need for Additional Advisory Consmittees Enrit GC LI (NOTE 3) 12/31/86 NA V.C.3 Study the IIeed to Estabitsh an Independent Nuclear Enrlt GC LI (NOTE 3) 12/31/86 NA Safety Board M

LICENSING PROCESS V.D.1 leprove Public and Intervenor Participation in the Eartt GC L1 (IIOTE 3) 12/31/86 NA Hearing Process V D.2 Study Construction-During-Adjudication Rules Enrit GC LI (NOTE 5) 12/31/86 NA V.D.3 Reexamine Commission Role in Adjudication Enrit GC LI (NOTE 5) 12/31/86 NA V.D.4 Study the Reform of the Licensing Process Emrit GC LI (NOTE 5) 12/31/86 NA M

LE6ISLATIVE IEEDS V.E.1 Study the Need for TNI-Related Legislation Enrit GC LI (NOTE 5) 12/31/86 NA g

ORGANIZATION AllD MANAGENENT t

V.F.1 Study NRC Top Management Structure and Process Emrit GC LI (NOTE 3) 12/31/86 NA V.F.2 Peexamine Organization and Functions of the NRC Offices Eerit GC LI (IIOTE 3) 12/31/86 NA V.F.3 Revise Delegations of Authority to Staff Enrtt GC LI (NOTE 3) 12/31/86 NA V.F.4 Clartfy and Strengthen the Respective Roles of Chairman, Eerit GC LI (NOTE 3) 12/31/86 NA i

Conestss1on, and Executive Director for Operations V.F.5 Authority to Delegate Emergency Response Functions Eerit GC LI (NOTE 3) 12/31/86 NA to a Single Comunissioner y

CONSOLIDATION OF NRC LOCATIONS v.G.1 Achieve Single location, Long-Tern Emrit GC LI (NOTE 3) 12/31/86 NA V.G.2 Achieve Single Location, Interim Eartt GC L1 (NOTE 3) 12/31/86 NA

o d

TASK ACTION PtAN ITEMS w$

A-1 Water Hanumer (former USI)

Enrtt NRR/ DST /GIB NOTE 3(a) 1 06/30/85 NA 8

A-2 Asynumetric Blowdown loads on Reactor Primary Coolant Emrit NRR/ DST /GIS NOTE 3(a) 1 06/30/85 D-10 o

O 3

m Systems (former USI)

W A-3 Westinghouse Steam Generator Tube Integrity (forimer USI)

Eerit NRR/ DEST /ENT8 NOTE 3(a) 1 12/31/88 ro

~

A-4 CE Steam Generator Tube Integrity (former USI)

Enrit NRR/ DES 1/ENTB NOTE 3(a) 1 12/31/88

Table II (Continued 1 Action Lead Office /

Safety Latest om Plan Iten/

Priority Division /

Priority latest Issuance MPA D Issue No.

T*tle Engineer Branch Ranking Revis'on Date No.

?.

m A-5 8W Steam Generator Tube Integrity (fomer USI)

Eartt NRR/ DEST /EMTB NOTE 3(a) 1 12/31/88 A-6 Mark I Short-Term Program (former USI)

Eartt NRR/ DST /GIB NOTE 3(a) 1 06/30/85 A-7 Mark I Long-Tem Program (former USI)

Eartt NRR/ DST /GIB NOTE 3(a) 1 06/30/85 D-01 A-8 Mark II Containment Pool Dyanmic Loads Long-Tern Emrit NRR/ DST /GIB NOTE 3(a) 1 06/30/85 NA Program (former USI)

A-9 ATVS (former USI)

Enrit NRR/ DST /GIS NOTE 3(a) 1 06/30/85 A-10 BWt Feedsater Nozzle Cracking (fomer US1)

Enrit NRR/ DST /GIB NOTE 3(a) 1 06/30/85 B-25 A-11 Reactor Vessel Materials Toughness (former USI)

Enrit NRR/ DST /GIB NOTE 3(a) 1 06/30/85 A-12 Fracture Toughness of Steam Generator and Reactor Emrit NRR/ DST /GIB NOTE 3(a) 1 06/30/85 NA Coolant Ptsup Supports (fomer USI)

A-13 Snubber Operabtitty Assurance Eartt NRR/DE/ME8 NOTE 3(a) 1 06/30/91 A-14 Flaw Detection Matthews NRR/DE/NTEB DROP 11/30/83 NA A-15 Primary Coolant System Decontamination and Steam Pittman NRR/DE/CHEB NOTE 3(b) 11/30/83 NA Generator Chemical Cleaning A-16 Steam Effects on BWR Core Spray Distribution Emrtt NRR/DSI/CFB NOTE 3(a) 11/30/83 D-12 A-17 Systems Interactions in Nuclear Power Plants (former Enrtt RES/DSIR/EIB NOTE 3(b) 1 12/31/89 NA (USI)

A-18 Pipe Rupture Design Criteria Enrtt NRR/DE/ME8 DROP 11/30/83 NA A-19 Digital Computer Protection System Mtistead RES/DSR/tFS LI (NOTE 5) 1 06/30/91 NA A-20 Impacts of the Coal Fuel Cycle NRR/DE/EHEB LI (NOTE 5) 11/30/83 NA

,A-21 Main Steamline Break Inside Containment - Evaluation of V*Molen NRR/DSI/CSB LOW 11/30/83 NA m

Environmental Conditions for Equipment Qualtf tcation A-22 PWR Main Steamline Break - Core, Reactor Vessel and V'Molen NRR/DSI/CSB DROP 11/30/83 NA Containment But1 ding Response A-23 Containment Leak Testing Matthews NRR/DSI/ CSS RI (NOTE Si 11/30/83 A-24 Qualificaticn of Class IE Safety-Related Equipment Enrtt NRR/ DST /GIS N0fE 3(a) 1 06/30/85 B-60 (fomer US!)

A-25 Non-Safety Loads on Class 1E Power Sources Thatcher NRR/DSI/PSB NOTE 3(a) 11/30/83 A-26 Reacter vessel Pressure Transtent Protection (former Eartt NRR/ DST /GIB NOTE 3(a) 1 06/30/85 8-04 (USI)

A-27 Reload Appitcations NRR/DSI/CPB LI (NOTE 5) 11/30/83 NA A-28 Increase in Spent Fuel Pool Storage Capacity Colmar NRR/DE/SGEB NOTE 3(a) 11/30/83 A-29 Nuclear Power Plant Design for the Redc tion of Colmar RES/DRPS/RPSI NOTE 3(b) 1 12/31/89 NA Vulnerabtitty to Industrial Sabotage A-30 Adequacy of Safety-Related DC Power Supplles Sege NRR/DSI/PSB 128 1

12/31/86 NA A-31 Rift Shutdown Requirements (former USI)

Enrit NRR/ DST /GIB NOTE 3(a) 1 06/30/85 A-32 Missile Effects Pittman hRR/DE/MTEB A-37, A-38, 11/30/83 NA B-68 A-33 NEPA Review of Accident Risks NRR/DSI/AEB El(NOTE 3) 11/30/83 NA A-34 Instruments for Monitoring Radiation and Process V'Molen NRR/DSI/ICSB II.F.3 11/30/83 NA M

variables During Accidents h A-35 Adequacy of Offstte Power Systems Eerit NRR/DS!/PSB NOTE 3(a) 1 12/31/94 p A-36 Control of Heavy Loads Near Spent Fuel (fomer USI)

Eartt NRR/DSI/GI8 NOTE 3(a) 1 06/30/85 C-10. C-15 g o A-37 Turbine Misslies Pittman NER/DE/MTEB DROP 11/30/83 NA 8 A-38 Tornado Missiles Sege NRR/DS!/ASB LOW I

12/31/94 NA 9

O A-39 Determination of Safety Relief Valve Pool Dynamic Enrit NRR/ DST /GIB NOTE 3(a) 1 06/30/85 y

Loads and Temperature Limits (former USI)

[

O O

O

--.n_.

N i

Table 11 (Continued)

Action Lead Office /

Safety Latest Q Plan Item /

Priority Division /

Priority latest Issuance WA

% !ssue No.

Title Engineer Branch Ranking Revision Date No.

$ A-40 Seismic Design Criteria (former USI)

Eerit RES/DSIR/EIB NOTE 3(a) 1 12/31/89 NA A-41 Long-Tern Seismic Program Colmar NRR/DE/MEB NOTE 3(b) 1 12/31/84 is A-42 Pipe Cracks in Botilng Water Reactors (former USI)

Enrtt NRR/ DST /GIB NOTE 3(a) 1 06/30/85 B-05 A-43 Containment Emergency Sump Performance (former USI)

Emrit NRR/ DST /6!B NOTE 3(a) 1 12/31/87 A-44 Station Blackout (former USI)

Enrit RES/DRPS/RPSI NOTE 3(a) 1 06/30/88 A-45 Shutdoun Decay Heat Removal Requirements (former USI)

Eurit RES/DRPS/RPSI NOTE 3(b) 1 12/31/88 NA A-46 Seismic Qualification of Equipment in Operating Plants Enrit NRR/DSR0/EIB NOTE 3(a) 1 12/31/87 (former USI)

A-47 Safety Impilcations of Contrc1 Systems (former USI)

Enrtt RES/DSIR/E18 NOTE 3(a) 1 12/31/89 A-48 Hydrogen Control Measures and Effects of Hydrogen Enrit NRR/DSIR/SAIB NOTE 3(a) 1-06/30/89 Burns on Safety Equipment A-49 Pressertred Thermal Shock (former USI)

Emrit NRR/DSR0/RSIB NOTE 3(a) 1 12/31/87 A-21 8-1 Envirormental Technical Specifications NRR/DE/EHEB EI (NOTE 3) 11/30/83 NA B-2 Forecasting Electricity Demand NRR EI (NOTE 3) 11/30/83 NA B-3 Event Categortration NRR/DSI/RSB LI (NOTE 3) 11/30/83 NA B-4 ECCS Reitablitty Emrit NRR/DSI/RSB II.E.3.2 11/30/83 NA B-5 Ductility of Tuo-Way Slabs and Shells and Buckling Thatcher RES/DE/EIB NOTE 3(b) 1 06/30/88 NA Behavior of Steel Containments B-6 Loads. Load Cont >tnations. Stress Limits Pittman NRR/DSR0/EIB 119.1 12/31/87 NA B-7 Secondary Accident Consequence Modeling NRR/DSI/AEB LI (NOTE 3) 11/30/83 NA B-8 Locking Out of ECCS Power Operated Valves Riggs NRR/DSI/RSB DROP 1

12/31/94 NA B-9 Electrical Cable Penetrations of Contatrament Emrtt NRR/DSI/PSB NOTE 3(b) 11/30/83 NA B-10 Behavior of BWR Mark III Contatruments V'Molen NRR/DSI/CS8 NOTE 3(a) 1 12/31/84 NA B-11 Subcompartment Standard Problems NRR/DSI/CSB LI (NOTE 5) 11/30/83 NA B-12 Containment Cooling Requirements (Non-LOCA)

Emrtt NRR/DSI/CSB NOTE 3(b)

I 12/31/86 NA B-13 Marviken Test Data Evaluation NRR/DSI/CSB LI (NOTE 5) 11/30/83 NA B-14 Study of Hydrogen Mixing Capablitty in Contatrument Emrit NRR/ DST /GIB A-48 11/30/83 NA Post-LOCA B-15 CONTEMPT Computer Code Maintenance NRR/DSI/CSB LI (NOTE 3) 11/30/83 NA B-16 Protection Against Postulated Piping Failures in Fluid Emrit NRR/DE/MEB A-18 11/30/83 NA

.Systens Outside Containment B-17 Criteria for Safety-Related Operator Actions Milstead RES/ DST /CIHFB MEDIUM 2

12/31/86 B-18 Vortex Suppression Requirements for Containment Sumps Emrit NRR/ DST /GIS A-43 11/30/83 NA B-19 Thermal-Hydraulic Stability Colmar NDR/DSI/CPB NOTE 3(b) 06/30/85 NA B-20 Standard Problem Analysis RES/DAE/AMBR LI (NOTE 5) 11/30/83 B-21 Core Physics NRR/DS!/CPB L1 (NOTE 3) 11/30/83 NA B-22 LWR Fuel Eartt RES/DSIR/RPSIB DROP 2

06/30/95 NA B-23 LMFBR Fuel NRR/DSI/CP8 LI (NOTE 3) 11/30/83 NA B-24 Setsmic Qualification of Electrical and Mechanical Enarit NRR A-46 11/30/83 NA Equipment B-25 Piping Benctumark Problems NRR/DE/MEB LI (NOTE 5) 11/30/83 x

y B-2C Structural Integrity of Containment Penetrations Riggs NRR/DE/MTEB NOTE 3(b) 1 12/31/84 NA m B-27 Implementation and Use of Subsection NF NRR/DE/MEB LI (NOTE 5) 11/30/83 i

Q B-28 Radionucitde/ Sediment Transport Program NRR/DE/EHEB El (NOTE 3) 11/30/83 NA a

B-29 Effectiveness of Ultimate Heat Sinks Pittman NRR/DE/EHEB LI (NOTE 3) 1 06/30/91 NA o

3 i

$ B-30 Design Basis Floods and Probability NRR/DE/EMEB LI (NOTE 5) 11/30/83 w B-31 Das Failure Model Milstead NRR/DE/SGEB LI (NOTE 3) 1 06/30/89 NA ro W B-32 Ice Effects on Safety-Related Water Suppites Pittman NRR/DE/EHEB 153 1

06/30/91 NA

[

l Table Il (Continued)

Actien g

Lead Office /

Safety Latest m Plan Item /

Priority Division /

Priority latest Issuance NPA D Issue No.

Title Engineer Branch Ranking Revision Date No.

k so m

8-33 Dose Assessment Methodology NRR/DSI/RAB LI (NOTE 3) 11/30/83 NA 8-34 Occupational Radiation Exposure Reduction Emrit NRR/DSI/RAB III.D.3.1 11/30/83 NA 8-35 Confirmation of Appendix I Models for Calculations of NRR/DSI/METB LI (NOTE 5) 11/30/83 Releases of Radioactive Materials in Gaseous and Liquid Effluents frcum Light Water Cooled Power Reactors B-36 Develop Design. Testing, and Maintenance Criteria for Emrit NRR/DSI/METB NOTE 3(a) 11/30/83 Atmosphere Cleanup System Air Filtration and Adsorption Units for Engineered Safety Feature Systems and for Nomal Ventilation Systems B-37 Chemical Discharges to Receiving Waters NRR/DE/EHEB EI (NOTE 5) 11/30/83 B-38 Reconnaissance Level Investigations NRR/DE/EHE8 El (NOTE 3) 11/30/83 NA 8-39 Transmission Lines NRR/DE/EHEB El (NOTE 3) 11/30/83 NA 8-40 Effects of Power Plant Entrainment on Plankton NRR/DE/EHE8 EI (NOTE 3) 11/30/83 NA 8-41 lapacts on Fisheries NRR/DE/EHEB El (NOTE 3) 11/30/83 NA 8-42 Socioeconomic Envirorunental Impacts NRR/DE/SAB El (NOTE 3) 11/30/83 NA B-43 Value of Aerial Photographs for Site Evaluation NRR/DE/EHE8 El (NOTE 5) 11/30/83 B-44 Forecasts of Generatin;i Costs of Coal and Nuclear Plants NRR/DE/SA8 EI (NOTE 3) 11/30/83 NA B-45 Need for Power - Energy Conservation NRR/DE/SAB El (NOTE 3) 11/30/83 NA B-46 Cost of Alternatives in Environmental Design NRR/DE/SAB EI (NOTE 3) 11/30/83 NA B-47 Inservice Insper.un of Supports-Classes 1, 2, 3, and Colmar NRR/DE/MTES DROP 11/30/83 NA MC Components oo B-48 BWR Control Rod Drive Mechanical Failures Enrit NRR/DE/MTE8 NOTE 3(b) 11/30/83 E-49 Inservice Inspection Criteria and Corrosion Prevention NRR L1 (NOTE 5) 11/30/83 Criteria for Contatruments 8-50 Post-Operating Basis Earthquake Inspection Colmar NRR/DE/SGE8 RI (NOTE 3) 1 06/30/85 NA B-51 Assessment of Inelastic Analysts Techniques for Emrit NRR/DE/MEB A-40 11/30/83 NA Equtpment and Copponents B-52 Fuel Assembly Seismic and LOCA Responses Emrtt NRR/ DST /GIS A-2 11/30/83 NA B-53 Load Break Saltch Sege NRR/DSI/PSB RI (NOTE 3) 11/30/83 B-54 Ice Condenser Contatruments Mtistead NRR/DSI/CSB NOTE 3(b) 1 12/31/84 NA 8-55 Improved Reliablitty of Target Rock Safety Relief V*Molen NRR/DE/EMEB MEDIlm 11/30/83 Valves 8-56 Diesel Reliability Mtistead RES/DRPS/RPSI NOTE 3(a) 2 06/30/95 0-19 8-57 Station Blackout Emrit NRR/ DST /GIB A-44 11/30/83 B-58 Passive Mechanical Failures Colmar NRR/DE/E08 NOTE 3(b) 1 12/31/85 NA B-59 (N-l) Loop Operation in BWRs and PWRs Colmar NRR/DSI/RS8 RI (NOTE 3) 1 06/30/85 E-04,E-05 B-60 Loose Parts Monitoring Systems Eerit NRR/DSI/CPB NOTE 3(b)

I 12/31/84 NA B-61 Allowable ECCS Equipment Outage Periods Pittman RES/ DST /PRAB MEDIUM 11/30/83 8-62 Reaxamination of Technical Bases for Establishing SLs, NRR/DSI/CFB LI (NOTE 3) 11/30/83 NA LSSSs, and Reactor Protection System Trip Functions B-63 Isolation of Low Pressure Systems Connected to the Emrtt NRR/DE/MEB NOTE 3(a) 11/30/83 y

2c Reactor Coolant Pressure Boundary 8-64 Deconentssioning of Reactors Colmar RES/DE/MEB NOTE 3(a) 2 06/30/95 NA g

c3 B-65 lodine Spiking Milstead NRR/DSI/AEB DROP 2

12/31/84 NA 6

B-66 Control Room Infiltration Measurements

$tatthews NRR/DSI/AEB NOTE 3(a) 11/30/83 0

3 e

B-67 Effluent and Process Monitoring Instrumentation Colmar NRR/DSI/METB III.D.2.1 11/30/83 NA y

B-68 Pump Dwerspeed During LOCA Riant NRR/DSI/ASB DROP 11/30/83 NA Fj B-69 ECCS Leakage Ex-Contattunent Riani NRR/DSI/t*ETB III.D.I.l(l) 11/30/83 NA m

O O

O

Table antinuedi Action Lead Office /

Safety latest

$ Plan item /

Priority Division /

Priority latest lasuance WA i

g Issue No.

Title Engineer Branch Ranking Revision Date No.

?

I

$ B-70 Power Grid Frequency Degradation and Effect on Primary Enrit NRR/DSI/PSB NOTE 3(b) 11/30/83 Coolant Pumps B-71 Incident Response Riant MRR III.A.3.1 11/30/83 lea 8-72 Health Effects and Life Shortening from Urantian and NRR/DSI/RA8 LI (NOTE 5)

!!/30/83 NA Coal Fuel Cycles t

8-73 Monitoring for Excessive Vibration Inside the Reactor Thatcher NRR/DE/MEB C-12 11/30/83 NA f

Pressure Vessel C-1 Assurance of Continuous Long ferm Capablitty of Hermetic Mtistead NRR/DE/EQB NOTE 3(a) 11/30/83 Seals on Instrismantation and Electrical Equipment C-2 Study of Contatrument Depressurization by Inadvertent Emrit NRR/DSI/CSB NOTE 3(b) 11/30/83 NA Spray Operation to Determine Adequacy of Containment External Design Pressure C-3 Insulation Usage Within Contaltunent Emrit NRR/ DST /6IB A-43 1

06/30/91 NA i

C-4 Statistical Methods for ECCS Analysis Riggs NRR/DSR0/SPEB RI (NOTE 3) 1 06/30/86 NA C-5 Decay Heat Update Riggs NRR/DSR0/SPEB RI (NOTE 3) 1 06/30/86 NA C-6 LOCA Heat Sources Riggs NRR/DSRO/SPEB RI (NOTE 3) 1 06/30/86 NA C-7 PWI System Piping Enrit NRR/DE/MTEB NOTE 3(b) 11/30/83 NA C-8 Main Steam Line Leakage Control Systens Mtistead RES/0RPS/RPSI NOTE 3(n) 1 06/30/90 NA C-9 RIGt Heat Exchanger Tube Failures V'Molen NRR/DSI/RS8 DROP 11/30/83 NA C-10 Effective Operation of Containment Sprays in a LOCA Emrit NRR/DSI/AE8 NOTE 3(a) 11/30/83 NA C-Il Assessment of Failure and Reliability of Pungs and Emrit NRR/DE/ME8 NOTE 3(b) 12/31/85 NA 1

Valves C-12 Primary System Vibration Assessment Thatcher NRR/DE/MEB NOTE 3(b) 11/30/83 NA C-13 Non-Randam Failures Emrit NRR/ DST /GIB A-17 1

06/30/91 NA C-14 Storm Surge Model for Coastal Sites Emrit NRR/DE/EHEB LI (NOTE 3) 06/30/88 NA C-15 NUREG Report for Liquid Tank Failure Analysis WRR/DE/EHE8 LI (NOTE 3) 11/30/83 NA C-16 Assessment of Agricultural Land in Relation to Power NRR/DE/EE B EI (NOTE 3) 11/50/83 NA Plant Siting and Cooling System Selection C-17 Interim Acceptance Criteria for Solidification Agents Emrit NRR/DSI/ MET 8 NOTE 3(a) 11/30/83 NA for Radioactive Solid Wastes D-1 Advisability of a Seismic Scram Thatcher RES/DET/MSEB LOW 11/30/83 NA D-2 Emergency Core Cooling System capability for Future Emrtt RES/DRA/ARGIB DROP 12/31/88 NA Plants D-3 Control Rod Drop Accident Emrit NRR/DSI/CPB NOTE 3(b) 11/30/83 NA NEW GENERIC ISSUES 1.

Failures in Air-Monitoring, Air-Cleaning, and Eerit.

MRR/DSI/METB DROP 11/30/83 NA Ventilating Systems 2.

Failure of Protective Devices on Essential Equipment Diab RES/DSIR/Elb DROP 2

06/30/95 NA x

3.

Set Point Drift in Instrumentation Emrit NRR/DSIR/RPS!B NOTE 3(b) 1 06/30/86 NA

'D x 4 End-of-Life and Maintenance Criteria Thatcher NRR/DE/EQB NOTE 3(b) 11/30/83 NA y

S.

Design Check and Audit of Balance-of-Plant Equipment Pittman NRR/DSI/ASB I.F.f 11/30/83 NA

{

a 6.

Separation of Control Rod from Its Drive and BWt High V'Molen NRR/DSI/CPB NOTE 3(b) 1 12/31/94 NA o

Rod Worth Events 3

w 7.

Failures Due to Flow-Induced Vibrations V*Molen NRR/DSI/RSB DROP 1

. D6/30/91 NA ru W

8.

Inadvertent Actuation of Safety Injection in PWRs Colmar NRR/DSI/RSS I.C.1 11/30/83 NA i

Table 11 (Continued)

Action Lead Office /

Safety latest

$ Plan item /

Priority Division /

Priority latest Issuance MPA g Issue No.

Title Engineer Branch Ranking Reviston Date No.

?$

9.

Reevaluation of Reactor Coolant Pts, Trip Criteria Emrit NRR/DSI/RSS II.K.3(5) 11/30/83 NA 10.

Surveillance and Maintenance of TIP lsolation Valves Riggs NRR/DSI/ICSB DROP 11/30/83 NA and Squib Charges 11.

Turbine Disc Cracking Pittman NRR/DE/MTEB A-37 11/30/83 NA 12.

BWR Jet Ptapp Integrity Sege NRR/DE/MTEB, NOTE 3(b) 1 12/31/84 NA MEB 13.

Small Break LDCA fram Extended Overheating of Riant NRR/DSI/RSB DRDP 11/30/83 NA Pressurizer Heaters 14.

PWR Pipe Cracks Emrit NRR/DE/MTEB NOTE 3(b) 2 12/31/94 NA 15.

Radiation Effects on Reactor Vessel Supports Emrit RES/DET/EMMEB NOTE 3(b) 3 06/30/96 NA 16.

DWR Main Steam Isolation Valve Leakage Control Systems Milstead NRR/DSI/ASB C-8 11/30/83 NA 17.

Loss of Offsite Power Subsequent to a LOCA Colmar NRR/E,51/PSB, DROP 11/30/83 NA ICSB 18.

Steam Line Break with Consequential Small LOCA Riggs NRR/DSI/RSS I.C.1 11/30/83 NA 19.

Safety Implications of Nonsafety Instrianent and Control Sege NRR/ DST /6IB A-47 11/30/83 NA Power Supply Eus 20.

Effects of Electranagnetic Pulse on Nuclear Power Thatcher NRR/DSI/ICSB NOTE 3(b) 1 06/30/84 NA Plants 21.

Vibration Qualification of Equipment Riggs NRR/DE/EIB DROP 2

06/30/91 NA 22.

Inadvertent Boron Dilution Events V'Malen NRR/DSI/R$8 NOTE 3(b) 2 12/31/94 NA 23.

Reactor Coolant Pisip Seal Failures Riggs RES/DET/6 SIB HIGH 11/30/83

$ 24.

Automatic ECCS Switchover to Recirculation Milstead RES/DET/6 SIB NOTE 3(b) 3 12/31/95 NA 25.

Automatte Air Header Dunp on BWR Scram System Milstead NRR/DSI/R$8 NOTE 3(a) 11/30/83 26.

Diesel Generator Loading Problems Related to SIS Reset Eerit NRR/DSI/ASB 17 11/30/83 NA on loss of Offsite Power 27.

Manual vs. Automated Actions Pittman NRR/DSI/RSB B-17 11/30/83 NA 28.

Pressurized Th-mal Shock Eerit NRR/ DST /618 A-49 11/30/83 NA 29.

Bolting Degradation or Failure in Nuclear Power Plants V Malen RES/DSIR/EIB NOTE 3(b) 2 06/30/95 NA 30.

Potential Generator Missiles - Generator Rotor Pittman NRR/DE/MEB DROP 1

12/31/85 NA Retaining Rings 31.

Natural Circulation Cooldown Riggs NRR/DSI/RSB I.C.1 11/30/83 NA 32.

Flow Blockage in Essential Equipment Caused by Corbicula Emrit NRR/DSI/ASB 51 11/30/83 NA 33.

Correcting Atmospheric Dism Valve Opening Upon loss of Pittman NRR/DS!/ICSB A-47 11/30/83 NA Integrated Control System Power 34.

RCS Leak Riggs NRR/DHFS/PSRB DRDP 1

06/30/84 NA 35.

Degradation of Internal Appurtenances in LWRs V'Molen NRR/DSI/CPB, LOW 06/30/85 NA R$8 36.

tess of Service Water Colmar NRR/DSI/ASB, NOTE 3(b) 3 06/30/91 NA AEB, RSB 37.

Steam Generator Overfill and Contined Priliary and Colmar NRR/ DST /6IB, A-47, 1

06/30/85 NA Secondary Blowdown NRR/DSI/R$8 1.C.l(2)

W

$ 38.

Potential Recirculation System Failure as a Consequence Eerit RES/DSIR/RPSIB DROP 2

06/30/95 NA m

of Ingestion of Containment Paint Flak?s or Other Fine Q

Debris s

39.

Potential for Unacceptable Interaction Between the CRD Pittman NRR/DSI/ASB 25 1

06/30/95 NA O"

System and Non-Essential Control Air System w 40.

Safety Concerns Associated with Pipe Breaks in the BWR Colmar NRR/DSI/ASB NOTE 3(a) 1 06/30/84 B-65 N

~

Scram System O

O O

l t

Table II fContinued)

Action Lead Office /

Safety Latest S Plan Itas/

Priority Division /

Priority Latest Issuance PPA g Issue No.

Title Engineer Branch Ranking Revision Date No.

?.

cn 41.

Om Scree Discharge Volume Systems V'Molen NRR/DSI/RSB NOTE 3(a) 11/30/83 8-58 42.

Combination Primary / Secondary System LOCA Riggs NRR/DSI/RSB

!.C.1 1

06/30/85 NA 43.

Reliability of Air Systems Niistead RLS/DSIR/RPSI NOTE 3(a) 2 12/31/88 44.

Failure of Saltmeter Cooling System N11 stead NRR/DSI/AS8 43 1

12/31/88 NA l

45.

Inaperability of Instrumentation Due to Extreme Cold Milstead NRR/DSI/ICSB NOTE 3(a) 2 06/30/91 Weather 46.

Loss of 125 Volt DC Bus Sege NRR/DSI/PS8 76 11/30/83 NA 47.

Loss of Offsite Power Thatcher NRR/DSI/RSB, NOTE 3(b) 11/30/83 ASB 48.

LCD for Class IE Vital Instrument Buses in Operating Sege NRR/USI/PSB 128 1

12/31/86 NA Reactors 49.

Interlocks and LCOs for Redundant Class IE Tle-Breakers Sege NPR/DSI/PSS 128 3

06/30/91 NA 50.

Reactor vessel Level Instrumentation in SWRs Thatcher NRR/DSI/RSB, NOTE 3(b) 1 12/31/84 NA ICSB 51.

Proposed Requirements fer Improving the Reitability of Enrit RES/DE/E!B NOTE 3(a) 1 12/31/89 Open Cycle Service Water Systems 52.

SSW Flow Blockage by Blue Mussels Emrit NRR/DSI/ASB 51 11/30/83 NA 53.

Consequences of a Postulated Flow Glockage Incident V'Molen NRR/DSI/CPB, DROP 1

12/31/84 NA in a SWR RSS 54.

Valve Operator-Related Events Occurring During 1978, Colmar NRR/DE/MEB II.E.6.1 1

06/30/85 NA 1979, and 1980 l

55.

Failure of Class IE Safety-Related Switchgear Circuit Enrit NRR/DSI/PSB DROP 2

06/30/91 NA l

Breakers to Close on Demand 56.

Abnormal Transient Dperating Guidelines as Applied to Colmar NRR/DHFS/HFEB A-47, 11/30/83 NA i

a Steam Generator Overft11 Event I.D.1 57.

Effects of Fire Protection Systen Actuation

- Mtistead RES/DRA/ARGIB NOTE 3(b) 3 06/30/95 NA L

on Safety-Related Equipment j

58.

Inadvertent Containment Flooding Sege NRR/DSI/ASB, DROP 11/30/83 CSB 59.

Technical Specification Requirements for Plant Shutdown Eerit NRR/ DST /TSIP RI (NOTE 5) 1 06/30/85 NA when Equipment for Safe Shutdown is Degraded or Inoperable 60.

Lamellar Tearing of Reactor Systems Structural Supports Colmar NRR/ DST /GIB A-12 11/30/83 NA 61.

SRV Line Break Inside the BWR Vetwell Airspace of Mark Nilstead NRR/DSI/CS8 NOTE 3(b) 2 12/31/86 NA I and Il Contairunents 62.

Reactor Systems Bolting Appilcations Riggs RES/DSIR/EIS 29 1

12/31/88 NA 63.

Use of Equipment Not Classified as Essential to Safety Pittman RES/DRA/ARGIB DROP 1

06/30/90 NA in BWR Transient Analysis 64.

Identification of Protection System Instrument Sensing Thatcher NRR/DSI/ICSB NOTE 3(b) 11/30/83 Lines M

65.

Probability of Core-Melt Due to Camponent Cooling Water V'Molen NRR/DSI/ASB 23 1

12/31/86 NA

'D Systen Failures E

66.

Steam Generator Requirements Riggs NRR/ DEST /ENTB NOTE 3(b) 2 12/31/88 NA O

8 67.

Steam Generator Staff Actions 8

67.2.1 Integrity of Steam Generator Tube Sleeves Riggs NRR/DE/ME8 135 4

06/30/94 NA g

67.3.1 Steam Generator Overfill Riggs NRR/ DST /GIB A-47, 4

06/30/94 NA

[

NRR/DSI/RSS

!.C.1 l

5 o

Table 11 (Continued)

Action Lead Office /

Safety latest

$ Plan Item /

Priority Division /

Priority Latest Issuance MPA N

lssue No.

Title Engineer Branch Ranking Revision Date No.

N$

67.3.2 Pressurtred Thermal Shock Riggs K2R/ DST /GIB A-49 4

06/30/94 NA 67.3.3 Isproved Accident Monitoring Riggs NRR/DSI/ICSB NOTE 3(a) 4 06/30/94 A-17 67.3.4 Reactor Vessel Inventory Measurement Riggs NRR/DSI/CPB II.F.2 4

06/30/94 NA 67.4.1 RCP Trip Riggs NRR/DSI/RSB II.K.3(5) 4 06/30/94 6-01 67.4.2 Control Room Design Review Riggs NRR/DHFS/HFEB I.D.1 4

06/30/94 F-08 67.4.3 Emergency Operating Procedures Riggs NRC/DHFS/PSRB I.C.1 4

06/30/94 F-05 67.5.1 Reassessment of Radiological Consequences Riggs RES/DRPS/RPSI L1 (NOTE 3) 4 06/30/94 NA 67.5.2 Reevaluation of $6TR Design Basis Riggs RES/DRPS/RPSI LI (67.5.1) 4 06/30/94 NA 67.5.3 Secondary System isolation Riggs NRR/DSI.1SB DROP 4

06/30/94 NA 67.6.0 Organizational Responses Riggs OIE/DEPER/lRDB III.A.3 4

06/30/94 NA 67.7.0 Isoroved Eddy Current Tests Riggs RES/DE/EIB 135 4

06/30/94 NA 67.8.0 Denting Criteria Riggs NRR/DE/MTEB 135 4

06/30/94 NA 67.9.0 Reactor Coolant System Pressure Control Riggs NRR/DSI/61B A-45, 4

06/30/94 NA NRR/DSI/R$8 I.C.1 (2,3) 67.10.0 Supplemental Tube Inspections Riggs NRR/DL/0RAB L1 (NOTE 5) 4 06/30/94 NA 68.

Postulated Loss of Auxiliary Feedwater System Resulting Pittman NRR/DSI/ASB 124 3

06/30/91 NA from Turbine-Driven Auxiliary Feedwater Pu m Steam Supply Line Rupture 69.

Make-up Mozzle Cracking in B&W Plants Colmar NRR/DE/MEB, NOTE 3(b) 1 12/31/84 B43 MTEB 70.

PORV and Block Valve Reltability Riggs RES/DE/EIB NOTE 3(a) 3 06/30/91 11.

Failure of Resin Demineraltrer Systems and Their Pittman RES/DRA/ARGIB LOW l

06/30/90 NA Effects on Nuclear Power Plant Safety 72.

Control Rod Drive Guide Tube Support Pin Failures Ri gg:-

RES DROP 1

06/30/91 NA 73.

Detached Thermal Sleeves Emrit RES/DS!R/EIB NOTE 3(a) 3 06/30/95 NA 74.

Reactor Coolant Activity Limits for Operating Reactors Milstead NRR/DSI/AEB DROP 1

06/30/86 NA 75.

Generic Iglications of ATVS Events at the Salem Emrit RES/DRA/ARGIB NOTE 3(a) 1 06/30/90 B-76,B-17 Nuclear Plant B-78.B-79 B-80,B-81 B-82 B-85 B-86,B-87 B-88.8-89 B-90,8-91 8-92,8-93 76.

Instrumentation and Control Power Interactions Zi merman RES/DSIR/EIB DROP 3

06/30/95 NA 77.

Flooding of Safety Equipment Ccacartments by Back-flow Colmar RFS/DE/E!B A-17 12/31/87 NA Through Floor Drains 78.

Monitoring of Fatigue Transler.. Limits for Reactor Rourk RES/DET/GSIB MEDIUM 2

06/30/95 Coolant System 19.

Unanalyzed Reactor Vessel Thermal Stress During Colmar RES/DSIR/EIB NOTE 3(b) 3 06/30/95 NA Natural Convection Cooldown M

80.

Pipe Break Effects on Control Rod Drive Hydraulle Lines V*Molen NRR/DSI/RSB, LOW I

06/30/91 NA k

y in the Drywells of BWR Mark I and 11 Containments

ASB, y

a CPB e

81.

Impact of Locked Doors and Barrters on Plant and Rourk RES/DSIR/EIB LOW 4

06/30/95 NA O

Personnel Safety w

82.

Beyond Design Basts Accidents in Spent Fuel Pools V*Molen RES/DRPS/RPSI NOTE 3(b) 1 06/30/89 NA N

83.

Control Room Habitability Emrtt RES/ DST /AEB NOTE 3(b) 2 06/30/96 NA O

O O

1 J

Table 11 (Continued 1 Action Lead Office /

Safety Latest

$ Plan Item /

Priority Division /

Priority Latest Issuance WA g Issue No.

Title Engineer Branch Rankirg Revision Date No.

?$

84.

CE PORVs Riggs RES/DSIR/RPSI NOTE 3(b) 2 06/30/90 NA 85.

Reliability of Vacuan Breakers Connected to Steam Milstead NRR/DSI/CSB DROP 2

06/30/91 NA Discharge Lines Inside BWt Contatruments 86.

Long Range Plan for Dealing with Stress Corrosion Emrit NRR/ DEST /ENTB NOTE 3(a) 1 06/30/88 8-84 6

Cracking in BWt Piping 87.

Failure of WCI Steam Line Without Isolation Pittman RES/DSIR/EIB NOTE 3(a) 2 06/30/95 88.

Earthquakes and Emergency Planning Riggs RES/DRA/ARGIB NOTE 3(b) 12/31/87 NA 89.

Stiff Pipe Clamps Chang RES/DSIR/EIB LOW 2

06/30/95 NA 90.

Technical Specifications for Anticipatory Trips V'Molen NRR/DSI/R$8, LOW I

06/30/95 NA ICSS 91.

Main Crankshaft Failures in Transamerica DeLaval Emrtt RES/DRA/ARGIB NOTE 3(b) 12/31!37 NA Emergency Diesel Generators 92.

Fuel Crum6 ling During LOCA V'Molen NRR/DSI/RSB, LOW 12/31/84 NA CPS 93.

Steam Binding of Auxiliary Feeduater Ptsups Pittman RES/DRPS/RPSI NOTE 3(a) 06/30/88 94.

Additional Low Temperature Overpressure Protection Pittman RES/DSIR/RPSI NOTE 3(a) 06/30/90 for Light Water Reactors 95.

Loss of Effective Voltsue for Contairment Recirculation Milstead RES/DRA/ARGIS NOTE 3(b) 06/30/90 NA Spray 96.

RMt Svetion Valve Testing Milstead RES/DRA/ARGIB 105 06/30/90 NA 97.

PWR Reactor Cavity thcontrolled Exposures V'Molen NRR/DSI/RA8 III.D.3.1 06/30/85 NA

$ 98.

CRD Accumulator Check Valve Leakage Pittman NRR/DSI/ASB DROP 06/30/85 NA 99.

RCS/RWt Suction Line Valve Interlock on PWRs Pittman RES/DRPS/RPSI NOTE 3(a) 3 06/30/91 100.

Once-Through Steam Generator Level Jackson RES/DSIR/EIB DROP 1

06/30/95 NA 101.

BWR Water Level Redundancy V'Molen RES/DE/EIB NOTE 3(b) 1 06/30/89 NA i

102.

Human Error in Everts involving wrong Unit or wrong Emrit NRR/DLPQ/LPEB NOTE 3(b) 2 12/31/88 NA Train 103.

Design for Probable Maxiansa Precipitation Emrit RES/DE/EIB NOTE 3(a) 1 12/31/89 NA 104 Reduction of Boron Dilutton Requirements Pittman RES/DRA/ARGIB DROP 12/31/88 NA 105.

Interfacing Systems LOCA at LWRs Nilstead RES/DE/EIB NOTE 3(b) 4 06/30/95 NA 106.

Piping and Use of Highly Com6ustible Gases in Vital Milstead RES/DRPS NOTE 3(b) 2 06/30/95 NA Areas 107.

Main Transformer Failures Nilstead RES/DRA/ARGIB LOW I

06/30/91 NA 108.

BWR Suppression Pool Temperature Limits Colmar NRR/DSI/CSB RI (NOTE 3) 06/30/85 NA 109.

Reactor Vessel Closure Failure Riggs RES/DRA/ARGIB DROP 06/30/90 NA 110.

Equipment Protective Devices on Engineered Safety Diab RES/DSIR/E18 DROP 1

06/30/95 NA Features 111.

Stress Corrosten Cracking of Pressure Bounda y Riggs NRR/DE/MTE8 LI (NOTE 5) 1 06/30/91 NA Ferritic Steels in Selected Environments 112.

W stinghouse RPS Surveillance Frequencies and Pittman NRR/DSI/ICSB RI (NOTE 3) 12/31/85 NA Out-of-Service Times N

$ 113.

Dynamic Qualtitcation Testing of Large Bore Riggs RES/DSIR/EIB NOTE 3(b) 2 06/30/95 NA k

M Hydraulle Snubbers

-^

$ 114.

Setsmic-Induced Relay Chatter Riggs NRR/DSR0/SPEB A-46 1

06/30/91 NA g 115.

Enhancement of the Reliability of Westinghouse Milstead RES/DRPS/RPSI NOTE 3(b) 06/30/89 NA y

e Solid State Protection System y 116.

Accident Management Pittman RES/DRA/ARGIB 5

06/30/91 NA

[

i e

Table II (Conttr:ued) action Lead Office /

Safety Latest oos Plan item /

Priority Division /

Priority latest Issuance MPA D Issue No.

Title Engineer Branch Ranking Revision Date No.

?

to O'

112.

Allowable Tise for Diverse Simultaneous Pittman RES/DRA/ARGIB DROP 06/30/90 NA Equipment outages 118.

Ten &n Anchorage Failure Shaukat RES/DSIR/EIB NOTE 3(a) 1 06/30/95 NA 119.

Plotna Review Committee Recaenendations 119.1 Piping Rupture Requirements and Decoupling of Riggs NRR/DE RI (NOTE 3) 2 06/30/93 NA SeIsaic and LOCA Loads 119.2 Piping Damping Values Riggs NRR/DE R1 (DROP) 2 06/30/93 NA 119.3 Decoupilng the OBE from the SSE Riggs NRR/DE RI (S) 2 06/30/93 NA 119.4 BWR Piping Materials Riggs NR4/DE RI (NOTE 5) 2 06/30/93 NA 119.5 Leak Detection Requirements Riggs NRR/DE RI (NOTE 5) 2 06/30/93 NA 120.

On-Line Testability of Protection Systems Milstead RES/DRA/ARGIB NOTE 3(b) 2 06/30/95 NA 121.

Hydrogen Control for targe, Dry PWR Containments Eartt RES/DSIR/SAIB NOTE 3(b) 2 06/30/95 NA 122.

Davis-Besse Loss of All Fee &ater Event of June 9. 1985: Short-Ters Actions 122.1 Potential Inability to Remove Reactor Decay Heat 122.1.a Failure of Isolation Valves in Closed Position V'Molen NRR/DSRO/RSIB 124 3

06/30/91 NA 122.1.b Recovery of Auxillary Feadwater V'Malen NRR/DSR0/RSIB 124 3

06/30/91 NA 122.1.c.

Interruption of Auxiliary Feedwater Flow V*Molen NRR/DSR0/RSIB 124 3

06/30/91 NA 122.2 Initiating Feed-and-Bleed V*Molen NRR/ DEST /SRIB NOTE 3(b) 3 06/30/91 NA 122.3 Physical Security System Constraints V*Molen NRR/DSR0/SPEB LOW 3

06/30/91 NA 123.

Deficiencies in the Regulations Governing DBA and Milstead RES/DSIR/SAIB DROP 1

06/30/95 NA m

.p.

Single-Failure Criteria Suggested by the Davis-Besse Event of June 9, 1985 124.

Auxtitary Feedwater System Reliability Emrit NRR/ DEST /SRXB NOTE 3(a) 3 06/30/91 125.

Davis-Besse Loss of All Feedwater Event of June 9. 1985: Lono-Tern Actions 125.I.1 Availablitty of the Shift Technical Advisor V'Malen PES /DRA/ARGIB DROP 6

12/31/89' NA 125.1.2 PORV Reliability 6

12/31/89 125.I.2.a Need for a Test Program to Estabitsh Reliability of V'Molen NRR/DSR0/SPEB 70 6

12/31/89 NA the PORY 125.I.2.b Need for PORV Surveillance Tests to Confirm V'Molen NRR/DSRO/SPEB 70 6

12/31/89 NA Operational Readiness 125.I.2.c Need for Additional Protection Against PORV Failure V*Molen NRR/DSR0/SPEB DROP 6

12/31/89 NA 125.I.2.d Capability of the PORV to Support Feed-and-Bleed V'Molen NRR/DSR0/SPEB A-45 6

12/31/89 NA 125.1.3 SPDS Availability Milstead RES/DRA/ARGIB NOTE 3(b) 6 12/31/89 NA 125.I.4 Plant-Specific Simulator Riggs RES/DRA/ARGIB DROP 6

12/31/89 NA 125.1.5 Safety Systems Tested in All Conditions Required by Riggs 9ES/DRA/ARGIB DROP 6

12/31/89 NA DBA 125.1.6 Valve Torque Limit and Bypass Switch Settings V'Molen RES/DRA/ARGIB DROP 6

12/31/89 NA 125.I.7 Operator Training Adeq-sacy 125.1.7.a Recover Failed Equipment Pittman RES/DRA/ARGIB DROP 6

12/31/89 NA y

2c 125.1.7.b Realistic Hands-On Training V*Molen RES/DRA/ARGIB DROP 6

12/31/89 NA 125.1.8 Procedures and Staffing for Reporting to NRC Emergency V'Molen RES/DRA/ARGIB DROP 6

12/31/89 NA 7

C)

Response Center 125.11.1 Need for Additional Actions on AFV Systems u>

125.II.l.a Two-Train AFW Unavailability V'Molen NRR/DSR0/SPEB DROP 6

12/31/89 NA 125.II.I.b Review Existing AFV Systems for Single Failure V*Molen NRR/DSRO/SPEB 124 6

12/31/89 NA 125.II.1.c NUREG-0737 eleliability Inprovements V'Molen NRR/DSR0/SPEB DROP 6

12/31/89 NA O

O O

m o

em

[

\\

d V

b Tf le II (Continued)

Action Lead Office /

Safety Latest

$ Plan Iten/

Priority Division /

Priority Latest Issuance MPA N Issue No.

Title Engineer Branch Ranking Revision Date No.

N

$ 125.II.I.d AFV/ Steam and Feedsater Rupture Control System /ICS V'Malen NRR/DSR0/SPE8 DROP 6

12/31/89 NA Interactions in B&W Plants 125.11.2 Adequacy of Existing Maintenance Requirements for Riggs RES/DRA/ARGIB DROP 6

12/31/89 NA Safety-Related Systee 125.II.3 Review Steam /Feedline Break Mitigation Systems for V'Molen NRR/DSR0/SPEB DROP 6

12/31/89 NA Single Fallure 125.II.4 Thermal Stress of OTSE Cogonents Riggs NRR/DSR0/SPEB DROP 6

12/31/89 NA 125.11.5 Theriaal-Hydraulic Effects of Loss and Restoration Riggs RES/DRA/ARGIB DROP 6

12/31/89 of Feedenter on Primarf System Components 125.II.6 Reexamine PRA Estimates of Core Danage Risk from Loss V *Molen RES/DRA/ARGIB DROP 6

12/31/89 NA of All Feeerster 125.11.7 Reevaluate Provision to Automatically Isolate V*Molen RES/DRPS/RPSI h0TE 3(b) 6 12/31/89 NA Feedwater from Steam Generator During a Line Break 125.11.8 Reassess Criteria for Feed-and-Bleed Initiation v'Molen RES/DRA/ARGI8 DROP 6

12/31/89 NA 125.II.9 Enhanced Feed-and-Bleed Capability V'Molen NRR/DSR0/SPEB DROP 6

12/31/89 NA 125.II.10 Hierarchy of Igrogtu Operator Actions Riggs RES/DRA/ARGIB DROP 6

12/31/89 NA 125.I1.11 Recovery of Main Fecesater as Alternative to Auxiliary Riggs RES/DRA/ARGIB DROP 6

12/31/89 NA Feedwater 125.II.12 Adequacy of Training R garding PORY Operation Riggs RES/DRA/ARGIB DROP 6

12/31/89 NA 125.II.13 Operator Job Aids Pittman NRR/DRA/ARGIB DROP 6

12/31/89 NA 125.11.14 Renote Operation of Equipmer.t Which Must Now Be V'Molen NRR/DSRO/SPEB L0ii 6

12/31/89 NA Operated Locally 126.

Reltabtitty of PWR Main Steam Safety Valves Riggs RES/DRA/ARGIB LI (NOTE 3) 06/30/88 NA 127.

Maintenance and Testing of Manual Valves in Safety-Pittman RES/DRA/ARGIB LOW 12/31/87 NA Related Systen's 128.

Electrical Power Reitability Emrit RES/DSIR/EIB NOTE 3(a) 2 06/30/95 129.

Valve Interlocks to Prevent Vessel Drainage During Milstead RES/DRA/ARGIB DROP 06/30/90 NA Shutdown Cooling 130.

Essential Service Water Pep Failures at Multiplant Riggs RES/DSIR/RPSIB NOTE 3(a) 2 12/31/95 Sites 131.

Potential Seismic Interaction Involving the Movable Riggs RES/DRA/ARGIB 5

1 06/30/91 NA In-Care Flux Mapping System Used in Westinghouse-Designed Plants 132.

R m System Inside Containment Su RES/DSIR/SAIB DROP 1

12/31/95 NA 133.

Update Policy Statement on Nuclear Plant Staff Pittman NRR/DLPQ/LHFB LI (NOTE 3) 12/31/91 NA Working Hours 134.

Rule on Degree and Erperience Requirement Pittman RES/DRA/RDB NOTE 3(b) 12/31/89 NA 135.

Steam Generator and Steam Line Overfill Emrit RES/DSIR/EIB NOTE 3(b) 3 06/30/95 NA 136.

Storage and Use of Large Quantities of Cryogenic Milstead RES/DRA/ARGIB LI (NOTE 3) 06/30/88 NA Con 6ustibles On Site 137.

Refueling Cavity Seal Failure Milstead RES/DRA/ARGIB DRDP 06/30/90 NA

o

$ 138.

Deinerting of BWR Mark I and II Containments During Milstead RES/DSIR/SAIB LOW I

06/30/95 NA

o Power Operations Upon Discovery of RCS Leakage or a Q

Train of a Safety System Inoperable I

e 139.

Thinning of Carbon Steel Piping in LWRs Riggs RES/DRA/ARGl8 RI (NOTE 3) 1 06/30/95 NA o

@ I40.

Fission ProdJct Removal Systems Riggs RES/DRA'ARGIB DROP 06/30/90 NA 141.

Large-Break LOCA With Consequential SGTR Riggs RES/DRA/ARGIS DROP 06/30/90 NA N

Table 11 (Continued)

Action Lead Office /

Safety Latest S Plan Itse/

Priority Dietston/

Priority latest Issuance IFA g Issue No.

Title Engineer Branen Renking Reviston Date No.

?W cn 142.

Leakage Through Electrical Isolators in M11 stead RES/DSIR/EIB NOTE 3(b) 3 06/30/95 NA Instrumentation Circuits 143.

Availability of Chilled Water Systems and Room Cooling Mtistead RES/DRA/ARGIB NOTE 3(b) 2 06/30/95 NA 144.

Scram Without a Turbine / Generator Trip Hrabal RES/DSIR/EIB LOW I

06/30/95 NA 145.

Actions to Rechsco Common Cause Failures Rasmuson RES/ DST /PRAB NOTE 1 1

06/30/95 146.

Support Flextbtitty of Equipment and Ccaponents Chang RES/DSIR/EIB NOTE 3(b) 2 06/30/95 NA 147.

Fire-Induced Alternate Shutdown / Control Room Panel Mtistead RES/DSIR/SAIB L1 (NOTE 3) 1 06/30/94 NA Interactions 148.

Smoke Control and Manual Fire-Fighting Effectiveness Basdekas RES/DSIR/RPSIB LI (NOTE 5) 12/31/92 NA 149.

Adequacy of Fire Barriers Eartt RES/DSIR/EIB LDW 1

06/30/95 NA 150.

Overpressurization of Cretainment Penetrations Mtistead RES/DSIR/SAIB DROP 1

06/30/95 NA 151.

Reltahtitty of !.nticipated Transtant Without Mtistead RES/DS!R/SAIB NOTE 3(b' 2

06/30/95 NA SCRAM Rectreulation Pius Trip in BWRs 152.

Design Basis for Valves That Might Be Subjected to Eartt RES/DSIR/EIB LOW 1

06/30/95 NA Significant Blowdown Loads 153.

Loss of Essential Service Water in LWRs Riggs RES/DRA/ARGIS NOTE 3(b) 2 12/31/95 NA 154.

Adequacy of Emergency and Essential Lighting Woods RES/DSIR/SAIB LOW I

06/30/95 NA 155.

Generic Concerns Aristna from TN1-2 CleLnts 155.1 More Realistic Source Tern Assiseptions Eerit RES/ DST /AEB NOTE 3(a) 2 06/30/95 NA 135.2 Estabitsh Licensing Requirements for Non-Operating Enrit RES/DSIR/EIB RI (NOTE 5) 2 06/30/95 NA Fact 11ttes

$ 155.3 leprove Design Requirements for Nuclear Factittles Earit RES/DSIR/ElB DROP 2

06/30/95 NA 155.4 Improve criticality Calculations Enrtt RES/DSIR/EIB DROP 2

06/30/95 NA 155.5 More Ree etic Severe Reactor Accident Scenario Eerit RES/DSIR/EIB DROP 2

06/30/95 NA 155.6 Inprove ce.:-ta=Ntton Regulations Emrit RES/DSIR/E18 DROP 2

06/30/95 NA 155.7 Improve De w ssioning Regulations Eerit RES/DSIR/EI8 DROP 2

06/30/95 NA 156.

Systematic E6eation Pecoram 156.1.1 Settlement of Foundationa and Buried Equipment Chang RES/DSIR/EIB DROP 4

06/30/95 NA 156.1.2 Dam Integrity and Site Flooding Chen RES/DSIR/SAIB DROP 4

06/30/95 NA 156.1.3 Site Hydrology and Ablitiy to Withstend Floods Chen RES/DSIR/SAIB DROP 4

06/30/95 NA 156.1.4 Industrial Hazards Ferrell RES/DSIR/SAIB DROP 4

06/30/95 NA 156.1.5 Tornado Missiles Chen RES/DSIR/SAIB DROP 4

06/30/95 NA 156.1.6 Turbine Missiles Enrit RES/DSIR/EIB DROP 4

06/30/95 NA 156.2.1 Severe Weather Effects on Structures Chen RES/DSIR/SAIB DROP 4

06/30/95 NA 156.2.2 Design Codes. Criteria, and Load Couttnations Kirkwood RES/DSIR/EIB DROP 4

06/30/95 NA 156.2.3 Containment Design and Inspection Shaukat RES/DSIR/EIB DROP 4

06/30/95 NA 156.2.4 Seismic Design of Structi:res, Systems, and Components Chen RES/DSIR/SAIS DROP 4

06/30/95 P4 156.3.1.1 Shutdown Systems Woods RES/DSIR/SAIS DROP 4

06/30/95 NA 156.3.1.2 Electrical Instrumentattui and Controls Woods RES/0!IR/SAIB DROP 4

06/30/9S NA 156.3.2 Service and Cooling Vatei Systeses Su RES/DSIR/SAIB DROP 4

06/30/95 NA 156.3.3 Venttiation Systems Burdick RES/DSIR/SAIB DROP 4

06/30/95 NA

,o 156.3.4 Isolation of High and Low Pressure Systems Burdick RES/DSIR/SAIB DROP 4

06/30/95 NA M 156.3.5 Automatic ECCS Switchover Milstead RES/DSIR/SAIB 24 4

C6/30/95 NA 156.3.6.1 Emergency AC Power Eerit RES/DSIR/EIB DR0p 4

06/30/95 NA f

156.3.6.2 Emergency DC Power Rourk ret /DSIR/EIB LOW 4

06/30/95 NA O

3

$ 156.3.8 Shared Systems Eartt RES/DSIR/EIB DRuP 4

06/30/95 NA w

156.4.1 RPS and ESFS Isolation Earlt RES/DSIR/EIB 142 4

06/30/95 NA ro 156.4.2 Testing of the RPS and ESFS Chang RES/DSIR/SAIB 120 4

06/30/95 KA O

O O

_m

-m c

O l

Table 11 (Continued)

Action O

Lead Office /

Safety Latest m Plan item /

Priority Division /

Priority Latest Issuance IFA D issue No.

Title Engineer Branch

' tanking Reviston Date 1o.

o.:. -

50 i

m 156.6.1 Pipe Break Effects on Systems and Components Page RES/DET/GSIB NOTE 4 4

06/3*/95 157.

Containment Performance Shaperow RES/DSIR/SAIB NOTE 3(b) 1 06/30/95 M

158.

Performance of Power-Operated Valves Under Design Hrabal RES/DET/GSIB KDItM 1

06/30/95 Basis Condttions

(

159.

Qualification of Safety-9 elated Pimsps While Running Su RES/DSIR/SAIB DROP 1

06/30/95 NA on Minlassa Flow I

160.

Spurious Actions of Instrumentation Upon Restoratton Rourk RES/DSIR/EIB DROP I

05/30/95 M

[

of Poewr

[

161.

Use of Non-Safety-Related Power Supplies in Safety-Rourk RES/DSIR/EIB DROP 1

06/30/95 NA i

Related Circuits

[

162.

Inadequate Tecimical Specifications for Shared Cheh RES/DSIR/SAIS DROP 1

06/30/95 M

t Systems at Italtiplant Sites When One tinit Is I

Shut Doun l

163.

Multiple Steam Generator Tube Leakage Burdick RES/DET/GSIB 310TE 4 (later) t 164.

Neutron Fluence in Reacter Vessel Eartt RES/DSIR/EIR DROP 1

06/30/95 M

165.

Safety and Safety / Relief Valve Reliability Hrabal RES/DET/GSIB HIGH 1

06/30/95 IIA l

166.

Adequacy of Fatigue Life of Metal Components Enrit NRR/DE/EE B NOTE I 1

06/30/95 r

167 Hydrogen Storage Factitty separation Burdick RES/DSIR/SAI8 LOW I

06/30/95 NA 168.

Environmental Qualtftcation of Electrical Equipment Enrtt NRR/DSSA/SPL8 NOTE 1 1

06/30/95 169.

BWR MSIV cm Mode Failure Due to Loss of Enrit RES/DET/GSIB NOTE 4 (later)

Accismulator Pressure 170.

Fuel Damage Criteria for H?gh Burnup Fuel Emrit RES/DET/GSIB NGTE 2 06/30/95 i

171.

E5F Failure fran LOOP Subsequent to a LOCA Rourk RES/DET/GSIB HIGH 06/30/95 172.

Multiple System Responses Program Eerit RES/DET/GSIB NOTE 2 12/31/95 173.

Scent Fuel Storace Pool 7

173.A Operating Factittles Eartt RES/DET/GSIB NOTE 2 12/31/95 f

173.8 Permanently Shutdown Facilities Emrit RES/DET/GSIB NOTE 2 12/31/95 174.

Fastener Gaetna Practices 174.A SONGS Employees' Concern Emrit RES/DET/GSIB NOTE 3(b) 12/31/95 174.8 Job son Gage Company Concern Enrtt RES/DET/GSIB NOTE 3(b) 12/31/95 t

175.

Nuclear Power Plant Shift Staffing Enrit RES/DET/GSIB NOTE 3(b) 12/31/95 176.

Loss of Fill-Oli in Rosemouat Transmitters Enrit RES/DET/GSIB NOTE 3(b) 12/31/95 177.

Vehicle Intrusion at TMI Enrit RES/DET/GSIB NOTE 3(a) 12/31/95 178.

Effect cir Hurricane Andrew on Turkey Point Emrtt RES/DET/GSIB L1 (NOTE 5) 12/31/95 179.

Core Performance Eerit RES/DET/GSIB LI (h0iE 5) 12/31/95 180.

Motice of Enforcement Discretion Enrtt RES/DET/GSIB LI (NOTE 3) 12/31/95 181.

Fire Protection Enrit RES/DET/GSIB LI (NOTE 5) 12/31/95 182.

General Electric Extended Power Uprate Enrit RES/DET/GSIB RI (NOTE 5) 12/31/95 183.

Cycle-Specific Parameter Limits in Technical Enrit RES/DET/GSIB RI (NOTE 5) 12/31/95 SpectfIcations 184.

Endangered Spectes Enrit RES/DET/GSIB EI (NOTE 5) 12/31/95 g

5 1

i HlRUI FACTURS ISSUES S

I o

O"$

HAFFINGAND004tlFICAT10NS ro HFt.1 Shift Staffing Pittman RES/DRPS/RHFB NOTE 3(a) 2 06/30/89

~

I

Table II (Continued 1 Action Lead Office /

Safety Latest

$ Plan item /

Priority Division /

Priority latest Issuance MPA N

Issue No.

Title Engineer Branch Ranking Revision Date No.

N$ El.2 Engineering Expertise on Shift Pittman NRR/DHFT/HFIB NOTE 3(b) 2 C6/30/89 W1.3 Guidance on Lianits and Conditions of Shif t Work Pittman NRR/DWT/WIB NOTE 3(b) 2 06/30/89 g

TRAINING W2.1 Evaluate Industry Training Pittman NRR/DWT/WIB LI (NOTE 5) 1 12/31/86 NA HF2.2 Evaluate INPO Accreditation Pittman NRR/DHFT/HFIB LI (NOTE 5)

I 12/31/86 NA W2.3 Revise SRP Section 13.2 Pittman NRR/DHFT/WIB LI (NOTE 5) 1 12/31/86 NA g

OPERATOR LICENSING EXAMINAff0NS HF3.1 Develop Job Knowledge Catalog Pittman NRR/DWT/WIB LI (NOTE 3) 2 12/31/87 NA HF3.2 Develop License Examination Handbook Pittman NRR/DWT/HFIB LI (NOTE 3) 2 12/31/87 NA HF3.3 Develop Criteria for Nuclear Power Plant Simulators Pittman NRR/DHFT/HFIB I.A.4.2(4) 2 12/31/87 NA HF3.4 Examination Requirements Pittman NRR/DWT/HFIB I.A.2.6(1) 2 12/31/87 NA HF3.5 Develos Ccuputertred Exam System Pittman NRR/DHFT/HFIB L1 (NOTE 3) 2 12/31/87 NA e

rR0ctDuRES HF4.1 Inspection Procedure for Dograded Emergency Pittman NRR/DLPQ/LHFB NOTE 3(b) 6 06/30/95 NA Operating Procedures

$ HF4.2 Procedures Generation Package Effectiveness Evaluation Pittman NRR/DHFT/WIB LI (NOTE 5) 6 06/30/95 NA HF4.3 Criteria for Safety-Related Operator Actions Pittman NRR/DHFT/HFIB B-17 6

06/30/95 NA HF4.4 Guidelines for Upgrading Other Procedures Pittman RES/DRPS/RHFB NOTE 3(b) 6 06/30/95 NA HF4.5 Application of Automation and Artificial Intelligence Pittman NRR/DHFT/HFIB HF5.2 6

06/30/95 NA Q

NAN 8tACHINE INTERFACE HF5.1 Local Control Stations Pittman RES/DRPS/RHF B NOTE 3(b) 4 06/30/95 NA HFS.2 Review Criteria for Human Factors Aspects of Advanced Pittman RES/DRPS/RHFB NOTE 3(b) 4 06/30/95 NA Controls and Instrumentation HF5.3 Evaluation of Operational Aid Systems Pittman NRR/DHFT/HFIB HF5.2 4

06/30/95 NA HF5.4 Ccmputers and Conputer Displays Pittman NRR/DHFT/HFIB 55.2 4

06/30/95 NA HF6 NANAGEMENT AND ORGANITATION HF6.1 Develop Regulatory Position on Management and Pittman NRR/DHFT/HFIB I.B.1.1 1

12/31/86 NA Organization (1.2.3.4)

HF6.2 Regulatory Position on Management and Organization Pittman NRR/DHFT/HFIB I.B.1.1 1

12/31/86 NA at Operating Reactors (1,2,3,4) x HlstAN REll APit1TY m

x

$ HFF.1 Htanan Error Data Acquisition Pittman NRR/DHFT/HFIB LI (NOTE 5)

I 12/31/86 NA E

I HF7.2 Hismen Error Data Storage arid Retrieval Pittman NRR/DHFT/HFIB LI (NOTE 5) 1 12/31/86 NA o

8 W7.3 Reliability Evaluation Specialist Aids Pittman NRR/DHFT/HFIB LI (NOTE 5) 1 12/31/86 NA g HF7.4 Safety Event Analysis Results Appilcations Pittman hRR/DHFT/HFIB L1 (NUTE 5) 1 12/31/86 NA

]

O O

O

'N N

(

\\

Table II (Continued)

Action Lead Office /

Safety Latest h Plan Itan/

Priority Division /

Priority Latest Issuance frA g Issue No.

Title Engineer Branch Ranking Revision Date No.

?.

m W8 Malatenance and Surveillance Program Pittman NRR/DLPQ/LPEB NOTE 3(b) 2 06/30/88 NA Cl(RNOSYL ISSE S 3

ADMINISTitATIVE CONTROLS MO OPERATIONAL PRACTICES CHl.1 Achinistrative Controls to Ensure That Procedures Are Follomed and That Procedures Are Adequate CHl.lA Symptom-Based E0Ps Emrit NRR/DLPQ/LHFB LI (NOTE 5) 06/30/89 NA CHl.lB Procedure Violations Emrit ItES/DSE/WRB LI (NOTE 5) 06/30/89 NA CHl.2 Approval of Tests and Other unusual Operations CHl.2A Test, Change, and Experiment Review Guidelines Enrit NRR/DOEA/DTSB LI (IIOTE 51 06/30/89 NA CHl.28 IstC Testing Requirements Emrit RES/DSR/lfRB LI (NOTE 5) 06/30/89 NA CHl.3 Bypassing Safety Systems CHl.3A Revise Regulatory Guide 1.41 Emrit RES/DE/EMEB LI (NOTE 5) 06/30/89 NA CHl.4 Availability of Engineered Safety Features CHl.4A Engineered Safety Feature Avallability Eerit NRR/DOEA/0TSB LI (NOTE 5) 06/30/89 NA CHl.45 Technical Specifications Bases Enrit IRR/DCEA/OTSB LI (NOTE 5) 06/30/85 NA CHl.4C Low Power and Shutdown Eerit RES/DSR/PRAB LI (NOTE 5) 06/30/89 NA L

CHl.5 Operating Staff Attitudes Toward Safety Emrit RES/DRA/ARGIB LI (NOTE 3) 06/30/89 NA E CHl.6 Management Systems CHl.64 Assessment of NRC Requirements on Management Enrit RES/DSR/HFRB LI (NOTE 5) 06/30/89 NA CHl.7 Accident Management Cht.7A Accident Management Emrit RES/DSR/HFRB LI (NOTE 5) 06/30/89 NA E

E CH2.1 Reactivity Accidents CH2.lA Reactivity Transients Eerit RES/DSR/RPSS LI (NOTE 5) 06/30/89 NA CH2.2 Accidents at Low Power and at Zero Power Emrit RES/DRA/ARGIB CHl.4 06/30/89 NA CH2.3 Miltiple-L%1t Protection CH2.3A Control Rom Habitability Enrit RES/DRA/ARGIB 83 06/30/89 NA CH2.38 Contamination Outside Contrni Room Eerit RES/DRA/ARGIB LI (NOTE 5) 06/30/89 NA CH2.3C Smoke Control Eerit RES/DSIR/SAIB LI (NOTE 5) 06/30/89 NA CH2.30 Shared Shutdown Systems Enrit RES/0RA/ARGIB LI (NOTE 5) 06/30/89 NA CH2.4 Fire Protection CH2.4A Firefighting With Radiation Present Emrit RES/DSIR/SAIS LI (IIOTE 5) 06/30/89 MA CONTAIIEENT b CH3.1 Containment Performance During Severe Accidents N CH3.lA Containment Performance Eerit RES/DSIR/SAIB LI (NOTE 5) 06/30/89 NA

$ CH3.2 Filtered venting E

g CH3.2A Flitered venting Emrit RES/DSIR/SAIB LI (NOTE 5) 06/30/89 NA g

U

Table II (Cecttrued)

Action Lead Office /

Safety Latest om Plan Item /

Priority Division /

Pricrity Latest Issuance WA N

Issue No.

Title Engineer Branch Ranking Revision Date No.

g to* M ENER6ENcv PLANNING CH4.1 Size of the Emergency Planning Zones Eerit RES/DRA/ARGIB LI (NOTE 3) 06/3afe; na CH4.2 Medical Services Emrit RES/DRA/ARGIB LI (NOTE 3) 06/30/89 NA CH4.3 Ingestion Pathway Measures CH4.3A Ingestion Pathway Protective Measures Eerit RES/DSIR/SA!B LI (NOTE 5) 06/30/89 NA CH4.4 De:cntamination and Relocation CH4.4A Decontamination Enrit RES/DSIR/5AIB LI (NOTE 5) 06/30/89 NA CH4.48 Relocation Enrit RES/DSIR/SAIB LI (NOTE 5) 06/30/89 NA M

SEVERE ACCIDENT PE N0E NA CH5.1 Source Tenn CH5.1A Mechanical Dispersal in Fission Product Release Emrit RES/DSR/AEB LI (NOTE 5) 06/30/89 NA CH5.lB Stripping in Fission Product Release Eerit RES/DSR/AEB LI (NOTE 5) 06/30/89 NA CHS.2 Steam Explosions CH5.2A 5 tease Explosions Emrtt RES/DSR/AEB LI (NOTE 5) 06/30/89 NA CH5.3 Combustible Gas Eerit RES/DRA/ARGIB L1 (NOTE 3) 06/30/89 NA M

GRAPHITE-MODERATED REACTORS CH6.1 Graphite-Moderated Reactors CH6.lA The Fort St. Vrain Reactor and the Modular HTGR Eerit RES/DRA/ARGIB LI (NOTE 3) 06/30/89 NA CH6.lB Structural Graphite Experiments Emrtt RES/DRA/ARGIB LI (NOTE 3) 06/30/89 NA CH6.2 Assessment Emrit RES/DRA/ARG1B LI (NOTE 3) 06/30/89 NA 2

Wi 5.

8 a

a 8

5 5

O O

O

s om N

i w

o t

N W

r m

i i

185L111 i

54f5WlY OF Tif PRIORITIZATIO5_9F f.L TMI ACTION PLAN ITElt5.

TASK ACTION PLAll ITENS. NEW GENERIC ISSKf. MGMll FACT (RS ISSUES. AND CER408YL ISSES Leeend l

ROTES:

I - Possible Resolution Identified for Evaluation 1

2 - Resolution Available 3 - Resolutions Resulted in either the Establishment of New Requirements or llo llew Requirements 4 - Issues to be Prioritized in the Future S - Issues that are r.ot GSIs but Should be Assigned Resources

{

for Completion OROP

- GSI Dropped from Further Pursuit EI

- Enviremental Issue GSI

- Generic Safety Issue i

HIGH

- High Safety Priority i

I

- TMI Action Plan Item with Implementation of Resolution Mendeted by NUREG-0737

[

LI

- Licensing Issue LOW

- Low Safety Priority l

EDItM

- Meditsu Safety Priority RI

- Regulatory Impact Issue l

USI

- Unresolved Safety Issue i

i L

[

I

o r

2 (D

t C

[

i llU o

o s

m w

I i

i

i Revision 21 O

W 2

R 2

R 2

=

m a

~

g

~

~

o wge 2

~

~

o o

g g

o 1

w g,

m m

o l

g e

i g

o m

i o

=

p 5

s o

m

~

o

=

W

_5 m

o i

o i

x o

o o

g

_v

-?

w 9

E m

N E

a w

G Y

g~

o aw 8C a

wg-o o

m o

m S

m a

g5gy g

e i

3 2

y, o

~

8 g

~

~

m 5

3 p

W E

W 2

~

m

=

=

g E

5 5

l r

t C

Y

=

u s

=

{g E

5 5

5 2

O h

5 E

~

C

[W 5 2

g 5,

~

~

9 W

5 s

E U

3 I

E E

m L

06/30/96 62 NUREG-0933

Revision 3 ISSUE 15: RADIATION EFFECTS ON REACTOR VESSEL SUPPORTS DESCRIPTION Historical Backaround This issue addressed the potential problem of radiation embrittlement of reactor i

vessel support structures (RVSS). It was originally identified as a Candidate USI in NUREG-0705" where it was recommended for further study before a judgment was made on its designation as a USI. In the initial prioritization of the issue in November 1983, it was concluded that the ORE associated with resolving the issue far outweighed the potential decrease in public risk. As a result, the issue was assigned a low priority until additional data on the problem became available 4

that would warrant a reevaluation of the issue. In April 1988, data developed by ORNL '"'" suggested that the potential embrittlement of the RVSS, as a result 2

i of neutron irradiation damage, could be significantly greater than was previously anticipated. Based on this new information, MEB/RES requested a reevaluation of

]

the issue in September 1988."

Neutron damage of structural materials causes embrittlement that may increase the potential for propagation of flaws that might exist in the materials. The potential for brittle fracture of these materials is typically measured in terms of the material's nil ductility transition temperature (NDTT), which is the lowest temperature at which the material would not be susceptible to failure by brittle fracture. As long as the operating environment in which the materials are 4

used has a higher temperature than the materials' NDTT, no failure by brittle fracture would be expected. Many materials, when subjected to neutron irradiation, experience an upward shift in the NDTT, i.e.,

they become more susceptible to brittle fracture at the operating temperatures of interest. This i

effect should have been accounted for in the design and fabrication of RVSS.

s However, the ORNL research indicated that the upward shift in NDTT with increased exposure to neutron irradiation had been underestimated. The loss in fracture toughness could result in failure of the RVSS and consequent movement of the i

reactor vessel, given the occurrence of a transient stress or shock such as would be experienced in an earthquake.

ORNL surveyed RVSS designs at LWRs and categorized each plant into one of five categories or types of RVSS: (1) skirt; (2) long-column; (3) shield-tank; 4

(4) short-column; and (5) suspension. Skirt supports are located away from the core with a large volume of intervening metal and water; radiation embrittlement of skirt RVSS was not anticipated. Long-column supports are located in a zone of potentially high neutron fluence and are thus susceptible to radiation damage.

Shield-tank supports are also located in a potentially high radiation damage zone. Short-column supports include several subcategories that are located in various regions relative to the reactor core and have a wide variability in i

susceptibility to radiation damage. Many plants with this type of support have special designs for heat dissipation, including natural convection, forced convection, and water / cooling-coil designs. Suspension supports were employed at

![

only one plant and, although these supports are located in a region of

\\

potentially high irradiation damage, the temperature may be high enough to preclude brittle fracture. However, for this analysis, plants employing the long-06/30/96 3.15-1 NUREG-0933

Revision 3 column, shield-tank, short-column, and suspension supports were assumed to be susceptible to irradiation damage.

Safety Sianificance A large seismic event can cause failure of auxiliary piping which can cause an embrittled RVSS to fracture thereby allowing the reactor vessel to move. Such movement can then worsen the resultant LOCA from the rupture of auxiliary piping by rupturing other piping attached to the primary coolant loop and instrument tubing attached to the bottom head of the reactor vessel.

Possible Solutions The proposed resolution for some plants involved the application of local heaters and insulation for the RVSS to maintain operating temperatures well above the NDTT of the potentially embrittled support. This resolution would only involve those plants that employed long-column and shield-tank supports. Short-column and suspension supports were in a higher temperature environment and thus heaters were not necessary to maintain the temperatures above the NDTT. However, minor design and equipment changes would be needed to control the amount of heat dissipation applied to the short-column and suspension supports to ensure the NDTT of the structural materials did not exceed the environmental temperature.

In all cases, appropriate safeguards must be installed to prevent overheating of the concrete around and in contact with the supports.

PRIORITY DETERMINATION Assumotigni The number of potentially susceptible plants (78) was determined from the results of the ORNL survey and are summarized below.

Number of Affected Plants Plant Tvoe RVSS Tvoe Operatina Under Construction PHR Short-column 45 13 Long-column 10 1

Shield-tank 8

0 Sub-Total :

63 14 BWR Suspension 1

0 Total:

64 14 The ORNL report also provided the basis for estimates of the length of time a plant could potentially operate in a vulnerable condition, i.e., with embrittled reactor vessel supports. The radiation embrittlement of RVSS materials froin two operating LWRs (Turkey Point and Trojan) was investigated and data on the change in NDTT over time were developed. The approximate time when the RVSS material was believed to become susceptible to brittle fracture occurred after 23 effective full power years. Therefore, the potential susceptibility of the RVSS to brittle fracture exists for 7 years at the end of a reactor's operating life, assuming an average operating life of 30 effective full power years. Data from the Oconee 06/30/96 3.15-2 NUREG-0933

Revision 3 O

3 and Grand Gulf 1 RSSMAP studies were used in this analysis to determine the Q

estimated risk for PWRs and BWRs, respectively.

Freauency Estimate The assumed accident scenario was the occurrence'of a seismic event of sufficient magnitude to cause fracture of an embrittled RVSS, subsequent movement of the reactor vessel, and a corresponding LOCA as attached piping ruptures. The analogous accident sequences were those involving LOCA initiators S, S,, and S 3

3 for Oconee (different initiator frequencies for three pipe diameters) and S for Grand Gulf; these are the corresponding LOCA initiators for pipe ruptures.

However, this issue was concerned with only seismically-induced pipe ruptures, which were not addressed in the original Oconee and Grand Gulf studies. As a result, seismically-induced LOCAs were defined here and incorporated into the l

base case.

The base case frequencies of seismically-induced LOCA initiators SS, SS,, SS,

i 3

and SS were assumed to be equal, i.e.,

the conditional probabilities of I

fracturing different sizes of pipe, given an earthquake, were assumed to be equal. Their base case frequencies were estimated as follows:

f(SS ) = f(SS,) = f(SS ) = f(S) = [f(PGA 2 0.29)][p(NDTT)][p(PR)]

3 3

where, f(PGA 2 0.2g) =

frequency of a seismic event with peak ground acceleration greater than or equal to 0.2g; frequency = 7 x 10/yr.

(O) p(NDTT) conditional probability that a RVSS is

=

susceptible to radiation damage and fails as a result of reactor vessel movement (this value is derived below).

p(PR) conditional probability of pipe rupture given movement of the reactor vessel

[ assumed to be accounted for in estimate of p(NOTT); effectively 1 for pipes of all diameters].

The conditional probability of failure of an embrittled RVSS as a result of a seismic event [p(NDTT)) is a function of the NDTT at the time the seismic event occurs, the number and size of preexisting flaws in the support material, and the safety factor built into the design of the supports and selection of the material. As discussed above, the RVSS materials at some plants may exceed operating temperatures during the last 7 years of reactor operation. Assuming that this occurs, the safety factor built into the RVSS may not exceed I whereas, using previous predictions of radiation damage, this safety factor may be as much determined'g a correlation" between safety factor and failure probability, PNL as 20. Usin that the conditional probability of failure leading to reactor core damage for a safety factor of 1 was 0.5. Using this value, the frequency of seismically-induced LOCAs was:

f(SS ) = f(SS,) = f(SS ) = f(S) = (7 x 10'*/,RY)(0.5)(1) p)

3 3

= 3.5 x 10' /RY xv 06/30/96 3.15-3 NUREG-0933

Revision 3 PNL derived the base case frequencies by substituting the above frequency of the seismically-induced initiators into the minimal cut sets given in NUREG/CR-2800. The results were as follows:

Oconee y(PWR-3)

- 1.4 x 10~'

SS H -

B (PWR-5) - 2.0 x 10-*/RY 3

e (PWR-7) - 1.4 x 10 */RY

/RY SS D -

a (PWR-1) = 2.4 x 10-'/RY 3

y (PWR-3) - 4.8 x 10-'

B (PWR-5) = 1.8 x 10-'/RY

/RY e (PWR-7) - 1.9 x 10-'/RY y (PWR-2) - 6.0 x 10-'

B (PWR-4) - 8.8 x 10-*/RY SS FH -

3 e (PWR-6) - 6.0 x 10-'/RY

/RY a (PWR-1) - 1.1 x 10-*

B (PWR-4) - 8.0 x 10-'/RY SS,FH -

e (PWR-6) - 8.8 x 10-'/RY

/RY SS,0 -

a (PWR-1) - 1.8 x 10-'/RY y (PWR-3) - 3.6 x 10 '

B (PWR-5) - 1.3 x 10-*/RY e (PWR-7) - 1.4 x 10-'/RY

/RY y (PWR-3) - 1.9 x 10 '

B (PWR-5) - 2.7 x 10-'/RY SS,0 -

/RY e (PWR-7) - 1.9 x 10-'/RY Grand Gulf a (BWR-1) - 1.2 x 10-'/RY S

3 6 (BWR-2) - 1.2 x 10 */RY Summing the base case frequencies for the affected release categories, the following were obtained:

Oconee Grand Gulf PWR 2.7 x 10-'/RY BWR 1.2 x 10-*/RY PWR 6.0 x 10-'

BWR 1.2 x 10-'/RY PWR 6.8 x 10-'/RY PWR-4 = 1.7 x 10^*/RY PWR 2.2 x 10-'/RY PWR 1.5 x 10^'/RY PWR 2.2 x 10-'/RY

/RY Based on the above, the affected base case core-melt frequency was 3.1 x 10 '/RY for PWRs and 1.2 x 10/RY for BWRs.

03/30/96 3.15-4 NUREG-0933

Revision 3 O

The possible solutions were assumed to eliminate the potential for radiation Q

embrittlement of RVSS materials. Thus, the adjusted case core-melt frequency was zero and the potential reduction in core-melt frequency was 3.1 x 10"/RY for PWRs and 1.2 x 10"/RY for BWRs.

Conseauence Estimate In order to obtain the consequences associated with this issue, the CRAC Code" was used. An average population density of 340 persons / square-mile was assumed (the average for U.S. domestic sites) from an exclusion area one-half mile about the reactor out to a 50-mile radius. A typical Midwest site meteorology was also assumed. Based on these assumptions, the risk for each Release Category was stated in Appendix D of NUREG/CR-2800." Using the frequency estimates derived above, the total estimated risk from the base case was 41.6 man-rem /RY for PWRs and 8.6 man-rem /RY for BWRS. Since the possible solutions were assumed to eliminate the potential for radiation embrittlement of RVSS materials, the adjusted case risk was essentially zero. The risk reduction associated with this issue was as follows:

EWR1: (41.6 man-rem /RY)(77 reactors)(7 years) - 22,400 man-rem ILE1: (8.6 man-rem /RY)(1 reactor)(7 years)

- 60 man-rem Therefore, the total potential risk reduction for all affected reactors was 22,460 man-rem.

I Cost Estimate

\\h Industry Cost: At operating plants, the solution consisted of controlling the temperature of the RVSS, either through application of local heaters and insulation or through controlling cooling systems that were already in place, to ensure that the temperatures of the structural materials did not fall below the materials' NDTT after irradiation embrittlement. At future plants, the use of non-susceptible materials was the proposed resolution. Since this could be accommodated during the design and construction stages of a plant, no additional costs were foreseen beyond those normally incurred during design and construction.

Affected backfit plants were assumed to implement the resolution after about ten years of reactor operation. It was further assumed that only plants with long-column and shield-tank supports would install and operate local heaters and insulation on their RVSS. The plants with suspension and short-column supports were assumed to implement measures to control or limit cooling of the RVSS.

Affected forward-fit plants would implement the solution before fuel was loaded i

into the core. The following was the breakdown of the solutions at the 78 affected plants:

PWRs:

(1) Backfit Heaters 18 Cooling Control 45 (2) Forward-fit 14

/

\\

V fLt:

Backfit (Cooling) 1 06/30/96 3.15-5 NUREG-0933

Revision 3 For plants with long-column and shield-tank supports, it was assumed that heaters would be attached to four reactor vessel support columns and that mounting hardware, metal-sheathed heating cables, switchgear, transformert, and a power l

controller would be installed. It was also assumed that the equipment would be installed during scheduled reactor outages. Therefore, no additional replacement power costs would be necessary. It was further assumed that access to the reactor cavity was possible for heater installation. PNL estimated the equipment cost to be $52,000/ plant; labor associated with installation of this equipment was estimated to be 105 man-weeks / plant. At a cost $2,270/ man-week, the installation cost for heaters was (105 man-weeks / plant)($2,270/ man-week) or $245,000/ plant.

An additional cost of $26,000/ plant was estimated for a Class V amendment.

Therefwe, the total implementation cost for plants that used heaters was

$320,000/ plant.

For plants with short-column and suspension supports that would utilize cooling methods, it was ussumed that equipment and labor requirements were 10% of that estimated for application of local heaters and insulation. In this case, PNL estimated the equipment cost to be $5,200/ plant; labor associated with installation of this equipment was estimated to be 10.5 man-weeks / plant. At a cost of $2,270/ man-week, the installation cost for cooling was (10.5 man-weeks / plant)($2,270/ man-week) or $25,000/ plant. The Cla: ? license amendment fee of $26,000/ plant would also be applicable. Therefore, the total implementation cost for plants that used cooling was $56,000/ plant.

Therefore, the total industry implementation cost was given by:

(18 p; ants)($320,000/ plant) + (46 plants)($56,000/ plant) - $8.34M l

PNL calculated that operation and maintenance costs would be $130,000/RY for those plants that use heaters and $7,100/RY for those that use cooling.

Therefore, the total operation and maintenance cost over the 7-year vulnerability period for t% affected reactors was given by:

(18 plants)(7 years){S130,000/RY) + (46 plants)(7 years)($7,100/RY) - $18.7M The total industry cost i r implementation, operation, and maintenance of the possible solutions was $(E 34 + 18.7)H or $27M.

NRC Cost: PNL estimated that it would require 16 man-weeks of staff effort to develop the possible solutions. At a rate of $2,270/ man-wsek, this amounted to

$36,000; contractor support was expected to cost an additional $500,000.

Therefore, the total NRC development cost was estimated to be $536,000.

NRC effort to support industry implementation of the solutions was estimated to be 15 man-weeks / plant for those with heaters and 2 man-weeks / plant for those with cooling. Assuming a rate of $2,270/ man-week, the total NRC implementation costs were:

$2,270[(18 plants)(15 man-wk/ plant) + (46 plants)(2 man-wk/ plant)] - $822,000 NRC review time for operation and maintenance was estimated to be 1 man-week /RY for all affected plants. At a cost of $2,270/ man-week, the total NRC cost for review of operation and maintenance of the possible solutions over the 7-year vulnerability period was given by:

06/30/96 3.15-6 NUREG-09L

i I

Revision 3 i

(64 plants)(7 years)($2,270/RY) - $1.02M i

Therefore, the total NRC cost for development., implementation, operation, and j

maintenance of the possible solutions was given by:

$(536,000 + 822,000 + 1,020,000) - $2.4M i

Total cost: The total industry and NRC cost associated with the possible i

solutions was $(27 + 2.4) or $29.4M.

Value/Imoact Assessment

)

3 i

Based on an estimated public risk reduction of 22,460 man-rem and a cost of

$29.4M for a possible solution, the value/ impact score was given by:

l' S - 22.460 man-rem

$29.4M l

762 man-rem /$M j

i Other Considerations No occupational dose would be incurred during implementation, operation, and maintenance of the solutions at forward-fit plants. Based on a radiation field i

of 100 millirem /hr in the vicinity of the reactor vessel, PNL estimated the total i

occupational dose increase of the 64 backfit plants to be 1,880 man-rem.

Operation and maintenance of the solutions at these plants were estimated to j

result in an additional risk of 5,100 man-rem. Thus, the total occupational dose i

increase from implementation, operation, and maintenance of the possible i

solutions was estimated to be 7,000 man-rem.

J i

Occupational dose reduction due to accident avoidance would be realized at the 1

forward-fit plants, as well as at backfit plants, over the last 7 years of I

reactor operation. The occupational dose reduction due to accident avoidance was j

calculated to be 330 man-rem for all 78 affected plants.

CONCLUSION l

Based on the potential public risk reduction and value/ impact score, the issue would have been given a medium priority ranking. Consideration of the net occupational dose increase associated with the possible solutions would not have i

changed this conclusion. However, because the change in core-melt frequency from i

implementation of the proposed solutions was estimated to be 3.1 x 10"/RY for

{

99% of the affected plants (PWRs), the issue was given a HIGH priority ranking.

4 Work completed by the staff in resolving the issue led to the preliminary conclusion that the potential problem did not pose an immediate threat to public health and safety; this conclusion was reported to the Commission in SECY 180" in June 1989.

Preliminary results from a new theoretical model for datermining damage by inw energy neutrons showed that the large NDTT shifts previously observed may have been a function of the particular neutron energy spectrum to which the steel samples were exposed. Data from the High Flux Isotope Reactor (HFIR) surveillance 06/30/96 3.15-7 NUREG-0933 j

Revision 3 program and from the neutron shield tank of the decommissioned Shippingport reactor were normalized to the same trend curve established from samples irradiated in materials test reactors. Extrapolation along the trend curve of NOTT change against exposure suggested that end-of-life values for RVSS would be on the order of one-third or one-fourth the value obtained by extrapolation of the HFIR data when plotted against traditional measures of neutron exposure.

Substantiation of these results would allow the staff to conclude that the predicted degradation in RVSS toughness (fracture resistance) would be insufficient to cause concern during the 40-year license life of each plant.

Additionally, preliminary results from the analysis of the Trojan plant indicated that RVSS failure would not cause failure of the reactor coolant system piping or reactor vessel in the event of an SSE or guillotine break in the pressurizer surge line. However, the ability of the control rods to scram and the integrity of instrument lines connected to the bottom of the reactor vessel under these conditions were to be confirmed. The Trojan RVSS was studied because of its configuration and certain significant design and fabrication details which the staff considered to be among the most vulnerable to failure under accident loads.

The above tentative results indicated that plant safety could be maintained despite RVSS radiation damage; however, extensive confirmatory analyses needed to be performed to support this preliminary conclusion.

The staff found that there was significant variability in the effect of radiation on RVSS among plants because of the variety of RPV support designs, material properties, and fuel management procedures that affected the neutron flux and spectrum in the cavity. In order to encompass the uncertainties in the various analyses and provide an overall conservative assessment, several structural analyses conducted demonstrated the following:

(1)

Postulating that one of four RPV supports was broken in a typical PWR, the remaining supports would carry the reactor vessel load even under SSE seismic loads; (2)

If all supports were assumed to be totally removed (i.e., broken),

the short span of piping between the vessel and the shield wall would support the load of the vessel.

The results of the analyses virtually eliminated the concern for both radiation embrittlement and significant structural damage from a postulated RPV failure.

A study of the neutron spectra at different HFIR pressure vessel surveillance locations and the staff's technical findings were published in NUREG/CR-6117"'

and NUREG-1509,* respectively. Based on the staff's regulatory analysis,"' the issue was RESOLVED with no new requirements."' Consideration of a license renewal period of 20 years did not change this conclusion.

REFERENCES 16.

WASH-1400 (NUREG-75/014), " Reactor Safety Study, An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," U.S. Nuclear Regulatory Commission, October 1975.

O 06/30/96 3.15-8 NUREG-0933

Revision 3 i O 44.

NUREG-0705, " Identification of New Unresolved Safety Issues Relating to i{V Nuclear Power Plant Stations," U.S. Nuclear Regulatory Commission, June 1981.

64.

NUREG/CR-2800,

" Guidelines for Nuclear Power Plant Safety Issue Prioritization Information Development,"

U.S.

Nuclear Regulatory Commission, February 1983, (Supplement 1) May 1983, (Supplement 2)

December 1983, (Supplement 3) September 1985, (Supplement 4) July 1986, (Supplement 5) July 1996.

408. NUREG-1509, " Radiation Effects on Reactor Pressure Vessel Support:;," U.S.

Nuclear Regulatory Commission, May 1996.

632.

NUREG/CR-6117, " Neutron Spectra at Different High Flux Isotope Reactor (HFIR) Pressure Vessel Surveillance Locations," U.S. Nuclear Regulatory Commission, December 1993.

672. Memorandum for J. Larkins from J. Murphy, " Proposed Resolution of GSI-15,

' Radiation Effects on Reactor Pressure Vessel Supports,'" June 22, 1994.

923. Memorandum for J. Taylor from D. Morrison, " Resolution of Generic Safety Issue 15, ' Radiation Effects on Reactor Vessel Supports,'" May 29, 1996.

1252. Memorandum for T.

King from C.

Serpan, " Reevaluation of Issue 15,

' Radiation Effects on Reactor Vessel Supports,'" September 30, 1988.

b 1253. ORNL/TM-10444, " Evaluation of HFIR Pressure-Vessel Integrity Considering Radiation Embrittlement," Oak Ridge National Laboratory, April 1988.

1254. NUREG/CR-5320, " Impact of Radiation Embrittlement on Integrity of Pressure Vessel Supports for Two PWR Plants," U.S. Nuclear Regulatory Commission, January 1989.

1255. UCLA-ENG-76113, "Some Probabilistic Aspects of the Seismic Risk of Nuclear Reactors," University of California, Los Angeles, December 1976.

1 1256. SECY-89-180, " Generic Safety Issue 15, ' Radiation Effects on Reactor Vessel Supports,'" June 13, 1989.

O 06/30/96 3.15-9 NUREG-0933

Revision 2 A

I (v

ISSUE 83: CONTROL ROOM HABITABILITY DESCRIPTION Historical Backaround On August 18, 1982, the ACRS issued a letter *" to the Commission which: (1) identified deficiencies in the maintenance and testing of engineered safety features designed to maintain control room habitability; (2) provided examples of design and installation errors, including inadvertent degradation of control room leak tightness; and (3) cited a shortage of NRC and licensee personnel knowledgeable about HVAC systems and nuclear air-cleaning technology. These ACRS concerns encompassed both plant licensing review and operations / inspection 4

activities. In January 1983, the staff responded'" to the ACRS concerns and recommended increased training of NRC and licensee personnel in inspection and testing of control room habitability systems. The staff also provided a profile of control room HVAC system component failures based on an analysis of LERs from 1977 through mid-1982. On April 28, 1983, NRR and 0IE representatives met with the ACRS Subcommittee on Reactor Radiological Effects to discuss the staff response.

In May 1983, the ACRS issued a letter'" to the ED0 which expressed continuing t ]/

concerns about control room habitability and provided both general and specific r

comments and recommendations for further staff evaluation. In July 1983, NRR D

transmitted to the EDO a joint NRR/0!E proposal'" for evaluating the ACRS comments and recommendations and the adequacy of the control room habitability licensing'" review process and criteria. In August 1983, the ED0 indicated agreement with the proposal and directed NRR to coordinate with OIE and the NRC Regional Offices to complete the program and submit a report to the ED0 by June 1,1984. In September 1983, NRR established'" a Control Room Habitability Working Group and a Steering Group for conducting and guiding the proposed review. Other generic issues that addressed related concerns were B-36, B-66, and III.D.3.4.

Safety Sianificance Loss of control rwa habitability following an accident release of external airborne toxic or radioactive material or smoke can impair or cause loss of the control room operators' capability to safely control the reactor and could lead to a core-damage accident. Use of the remote shutdown station outside the control room following such events is unreliable since this station has no emergency habitability or radiation protection provisicns similar to the control room.

Possible Solution The Control Room Habitability Work Group was expected to identify any recommended actions that would correct significant deficiencies in control room habitability design, installation, test, or maintenance.

J 06/30/96 3.83-1 NUREG-0933

Revision 2 CONCLUSION In June 1984, NRR provided a report to the ED0 along with its plans for implementing the recommendations of the report, including a survey of several operating plants. Based on the ongoing staff work, it was concluded that a solution had been identified and a schedule'" for the resolution of the issue was developed by DSI/NRR.

PNL completed a report'" entitled "A Probabilistic Examination of Nuclear Power Plant Control Room Habitability During Various Accident Scenarios," and the findings of the survey of operating plants were published in NUREG/CR-4960.""

As a result of these studies, it was recognized that the methodology used to evaluate control room habitability system design needed improvement. Accordingly, the staff initiated activities to develop: (1) improved methods for calculating control room dose and exposure levels; (2) improved meteorological models for use in control room habitability calculations; and (3) revised exposure limits to toxic gases for control room operators.

The results of the improved methods were documented in NUREG/CR-5669"' and NUREG/CR-6210"' and the HABIT Code was developed to provide an integrated code package for evaluating control room habitability. NUREG-1465,"" published with the resolution of Issue 155.1, will provide updated source term information for the evaluation of control room designs. As recommended *" by the ACRS, the staff was expected to consider NIOSH recommendations for toxic chemicals in its revision of Regulatory Guide 1.78."" Thus, this issue was RESOLVED with no new requirements.' Consideration of a license renewal period of 20 years would not change this conclusion.

REFERENCES 247. NUREG/CR-5669, " Evaluation of Exposure Limits to Toxic Gases for Nuclear Reactor Control Room Operators," U.S. Nuclear Regulatory Commission, July 1991.

249. NUREG/CR-6210, " Computer Codes for Evaluation of Control Room Habitability (HABIT)," U.S. Nuclear Regulatory Commission, June 1996.

296.

Letter to T. Kress from J. Taylor, " Resolution of Generic Safety Issue 83,

' Control Room Habitability,'" September 13, 1995.

335. Memorandum for J. Taylor from D. Morrison, " Resolution of Generic Safety Issue 83, ' Control Room Habitability,'" June 17, 1996.

436.

Letter to J. Ray from W. Dircks, " August 18, 1982, ACRS Letter on Control Room Habitability," January 31, 1983.

671.

Letter to N.

Palladino from P.

Shewmon, " Control Roorr Habitability,"

August 18, 1982.

673.

Letter to W. Dircks from J. Ebersole, "ACRS Subcommittee Report on Control Room Habitability," May 17, 1983.

674. Memorandum for W. Dircks from H. Denton, " Control Room Habitability," July 27, 1983.

06/30/96 3.83-2 NUREG-0933

i Revision 2 lO 675. Memorandum for H. Denton from W. Dircks, " Control Room Habitability,"

August 15, 1983.

676. Memorandum for T.

Murley, et al., from H.

Denton,

" Control Room Habitability," September 19, 1983.

677.

Letter to W.

Milstead (NRC) from T.

Powers (PNL),

"A Probabilistic Examination of Nuclear Power Plant Control Room Habitability During Various Accident Scenarios," December 3, 1984.

l 678. Memorandum for W. Dircks from H. Denton, " Control Room Habitability," June 29, 1984.

l 679. Memorandum for T. Speis from R. Bernero, " Revised Schedule for Generic Issue 83, Control Room Habitability," September 28, 1984.

1371. NUREG/CR-4960, " Control Room Habitability Survey of Licensed Commercial Nuclear Power Generating Stations," U.S. Nuclear Regulatory Commission, October 1988.

1373. Regulatory Guide 1.78, " Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release," U.S. Nuclear Regulatory Commission,. June 1974.

1465. NUREG-1465, " Accident Source Terms for Light-Water Nuclear Power Plants,"

U.S. Nuclear Regulatory Commission, February 1995.

l t

06/30/96 3.83-3 NUREG-0933

Revision 1 O

APPENDIX A 1

RELEASES FROM CONTAINMENT This appendix contains an excerpt from WASH-1400, Appendix VI (pp. 2-1 to 2-5),

that defined the radioactive Relea,a Categories upon which the Estimated Public Doses in Exhibit B in the Introduction were based.

i O

O 06/30/96 A.A-1 NUREG-0933

Revision 1 Section 2 Releases from Containment 2.1 CENERAL REMARKS A large portion of the work of the Reactor Safety Study was expended in determining the probability and magnitude of various radioactive releases. This work is described j

in detail in the preceding appendices as well as Appendices VII, and VIII.

In order to define the various releases that might occur, a series of release categories were identified for the postulated types of containment failure in both BWRs and PWRs. The probability of each release category and the associated magnitude of radioactive releases (as fractions of the initial core radioactivity that might leak from the containment structure) are used as input data to the consequence model.

In addition to probability and release magnitude, the parameters that characterize the various hypothetical accident sequences are time of relense, duration of release, warning time for evacuation, height of release, and energy content of the released plume.

The time of release refers to the time interval between the start of the hypothetical accident and the release of redioactive material from the containment building to the atmospherer it is used to calculate the initial decay of radioactivity. The duration of release is the total time during which radioactive material is emitted into the atmosphere it is used to account for continuous releases by adjusting for horizontal dispersion due to wind meander. These parameters, time and duration of release, represent the temporal behavior of the release in the dispersion model.

Th2y are used to model a " puff" release from the calculations of release versus time presented in Appendix V.

The warning time for evacuation (see section 11.1.1) is the interval between awareness of impending core melt and the release of radioactive material from the containment building. Finally, the height of release and the energy content of the released plume gas affect the manner in which the plume would be dispersed in the atmosphere.

Table VI 2-1 lists the leakage parameters that characterize the PWR and BWR release categories. It should be understood that these categories are composites of numerous event tree sequences with similar characteristics, as discussed in Appendix V.

2.2 ACCIDENT DESCRIPTIONS To help the reader understand the postulated containment releases, this section presents brief descriptions of the various physical proJesses that define each release category. For more detailed information on the release categories and the techniques employed to compute the radioactive releases to the atmosphere, the reader is referred to Appendices V, VII, and VIII. The dominant event tree sequences in each release category are discussed in detail in section 4.6 of Appendix V.

PWR 1 This release category can be characterized by a core meltdown followed by a steam explosion on contact of molten fuel with the residual water in the reactor vessel.

The containment spray and heat removal systems are also assumed to have failed and, therefore, the containment could be at a pressure above ambient at the time of the steam explosion. It is assumed that the steam explosion would rupture the upper portion of the reactor vessel and breach the containment barrier, with the result that a ethstantial amount of radioactivity might be released from the containment in a pufi over a period of about 10 minutes. Due to the sweeping action of gases generated during containment-vessel meltthrough, the release of radioactive materials would continue at a relatively low rate thereafter. The total release would contain 06/30/96 A.A-2 NUREG-0933

i i.

l Revision 1 i

j approximately 70% of the iodines and 40% of the alkali metals present in the core at the time of release.1 Because the containment would contain hot pressurized gases at the time of f ailure, a relatively high release rate of sensible energy from the containment could be associated with this category. This category also includes certain potential accident sequences that would involve the occurrence of core molting and a steam explosion af ter containment rupture dua to overpressure.

l In these sequences, the rate of energy release would be lower, although still relatively high.

PwR 2 This category is associated with the failure of core-cooling systems and oore melting concurrent with the failure of containment spray and heat-removal systems.

Failure of the containment barrier would occur through overpreissure, causing a substantial fraction of the containment atmosphere to be released in a puff over a period of about 30 minutes. Due to the sweeping action of gases generated during containment vessel meltthrough, the release of radioactive material would continue at a relatively low rate thereafter. The total release would contain approximately 70% of the iodines and 50% of the alkali metals present in the core at the time of release. As in PWR release category 1, the high temperature and pressure within containment at the time of containment failure would result in a relatively high release rate cf sensible energy from the containment.

PWR 3 This category involves an overpressure failure of the containment due to failure of containment heat removal. Containment failure would occur prior to the commencement of core molting. Core melting then would cause radioactive materials to be released through a ruptured containment barrier. Approximately 20% of the iodines and 200 of the alkali metals present in the core at the time of release would be released to the atmosphere. Most of the release would occur over a period of about 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The release of radioactive material from containment would be caused by the sweeping action of gases generated by the reaction of the molten fuel with concrete. Since these gases would be initially heated by contact with the melt, the rate of sensible energy release to the atmosphere would be moderately high.

PwR 4 This category involves failure of the core-cooling system and the containment spray injection system after a loss-of-coolant accident, together with a concurrent failure of the containment system to properly isolate. This would result in the release of 9% of the lodines and 44 of the alkali metals present in the core at the time of release. Most of the release would occur continuously over a period of 2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Because the containment recirculation spray and heat-removal systems would operate to remove heat from the containment atmosphere during core molting, a relatively low rate of release of sensible energy would be associated with this category.

Pwn 5 This category involves f ailure of the core cooling systems and is similar to pWR release category 4, except that the containment spray injection system would operate to further reduce the quantity of airborne radioactive material and to initially suppress containment temperature and pressure. The containment barrier would have a large leakage rate due to a concurrent failure of the containment system to properly isolate, and most of the radioactive material would be released continuously over a period of several hours. Approximately 34 of the iodines and 0.9% of the alkali metals present in the core would be released. Because of the operation of the containment heat-removal systems, the energy release rate would be low.

% e release fractions of all the chemical species are listed in Table VI 2-1.

The release fractions of iodine and alkali metals are indicated here to illustrate the variations in release with release category.

06/30/96 A.A-3 NUREG-0933

Revision 1 O

PwR 6 This category involves a core meltdown due to f ailure in the core cooling systems.

The containment sprays would not operate, but the containment barrier would retain its integrity until the molten core proceeded to melt through the concrete containment base mat.

The radioactive materials would be released into the ground, with some leakage to the atmosphere occurring upward through the ground. Direct leakage to the atmosphere would also occur at a low rate prior to containment-vessel meltthrough.

Most of the release would occur continuously over a period of about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

The release would include approximately 0.08% of the iodines and alkali metals present in the core at the time of release. Because leakage from containment to the atmosphere would be low and gases escaping through the ground would be cooled by contact with the soil, the energy release rate would be very low.

PwR 7 This category is similar to PWR release category 6, except that containment sprays would operate to reduce the containment temperature and pressure as well as the amount of airborne radioactivity. The release would involve 0.002% of the iodines and 0.001% of the alkali metals present in the core at the time of release. Most of the release would occur over a period of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. As in PWR release category 6, the energy release rate would be very low.

PwR 8 This category approximates a PWR design basis accident (large pipe break), except that the containment would fail to isolate properly on demand. The other engineered safeguards are assumed to function properly. The core would not melt. The release would involve approximately 0.01% of the iodines and 0.05% of the alkali metals.

Most of the release would occur in the 0.5-hour period during which containment pressure would be above ambient. Because containment sprays would operate and core molting would not occur, the energy release rate would also be low.

Pwn e This category approximates a PWR design basis accident (large pipe break) in which oniv the activity initially contained within the gap between the fuel pellet and claBding would be released into the containment. ne core would not molt. It is assumed that the minimum required engineered safeguards would function satisfactorily to remove heat from the core and containment. The release would occur over the 0.5-hour period during which the containment pressure would be above ambient.

Approximately 0.00001% of the iodines and 0.00006% of the alkali metals would be released. As in PWR release category 8, the energy release rate would be very low.

swa 1 This release category is representative of a core meltdown followed by a steam explosion in the reactor vessel. The latter would cause the release of a substantial quantity of radioactive material to the atmosphere. The total release would contain approximately 40% of the iodines and alkali metals present in the core at the time of containment failure. Most of the release would occur over a 1/2 hour period.

Because of the energy generated in the steam explosion, this category would be characterized by a relatively high rate of energy release to the atmosphere. This category also includes certain sequences that involve overpressure failure of the containment prior to the occurrence of core melting and a steam explosion. In these sequences, the rate of energy release would be somewhat smaller than for those discussed above, although it would still be relatively high.

O 06/30/96 A.A-4 NUREG-0933

i, Revision 1

!l&

sea 2 l

l This release category is representative of a core meltdown resulting from a transient event in which decay-heat-removal systems are assumed to fail. Containment over-pressure failure would result, and oore melting would follow. Most of the release 1

would occur over a period of about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The containment failure would be such i

that radioactivity would be released directly to the atmosphere without si,gnificant retention of fission products.

This category involves a relatively high rate of energy release due to the sweeping action of the gases generated by the molten mass.

Approximately 90% of the iodines and 50% of the alkali metals present in the core j

would be released to the atmosphere.

l sua 3 i

This release category represents a core meltdown caused by a transient event accompanied by a failure to scram or failure to remove decay heat. Containment failure would occur either before core melt or as a result of gases generated during the inter-t action of the molten fuel with concrete after reactor-vessel moltthrough. Some fission-product retention would occur either in the suppression pool or the reactor building prior to release to the atmosphere. Most of the release would occur ovar 2

a period of about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and would involve 10% of the iodines and lot of the alkali i

metals. For those sequences in which the containment would fail due to overpressure j

af ter core melt, the rate of energy release to the atmosphere would be relatively j

high. For those sequences in which overpressure failure would occur before core melt, the energy release rate would be somewhat smaller, although still moderately a

high.

i i

sNa 4 This release category is representative of a core meltdown with enough containment

'}

leakage to the reactor building to prevent containment failure by overpressure. The quantity of radioactivity released to the atmosphere would be significantly reduced by normal ventilation paths in the reactor building and potential mitigation by the escondary containment filter systems. Condensation in the containment and the action 1

of the standby gas treatment system on the releases would also lead to a low rate d

of energy release. The radioactive material would be released from the reactor l

building or the stack at an elevated level. Most of the release would occur over i

a 2-hour period and would involve approximately 0.08% of the iodines and 0.5% of the I

alkali metals.

a i

swa 5 This category approximates a BWR design basis accident (large pipe break) in which only the activity initially contained within the gap between the fuel pellet and cladding would be released into containment. The core would not molt, and containment leakage would be small. It is assumed that the minimum required engineered safe-l guards would function satisfactorily. The release would be filtered and pass through l

s the elevated stac.k.

It would occur over a period of about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> while the of the iodines and 4 x 10'7 containment is pressurised above ambient and would involve approximately 6 x 10" t t of the alkali metals. Since core melt would not occur

}

and containment heat-removal systems would c rate, the release to the atmosphere j

would involve a negligibly small amount of t ermal energy.

i i

i t

i 1

i l

4

\\

1 06/30/96 A.A-5 NUREG-0933

O C)

%W O

NW C4 TABLE VI 2-1

SUMMARY

OF RELEASE CATECORIES REPRESENTING HYP(YTHETICAL ACCIDENTS

"#*"M'I**"Y*'I'****m Ttme of eurataan marease Ttme slowet aan talemme 9eekeht 11ty Baaseos of metenes fut sweemtsee of me. ease gnargy holesse f---

i.. I thr) thet thr9 tueeeeet (106 Den /hrt me-as cr9amac I T

co-g>

go gb so-se autel R.a

-4 WW I 9 m RG"

2. S G.S R.e 25 2e eng S2e 0.9 6 s 16"I 4.7 e.4 c.4 e.oS e.4 3 e 30 9.9 1 e te O. 7

e. S

e. 3 e.06 6.02 4 a le"I

• I WW 2 S a 18"O 2.9 e.S 3.9 g

gio

-3 80 6 a te* E 9.2 0.2

9. k D.02 a.9 5 3 a 10 WWE S 4 a le S.O 1.S 2.9 9

6 4

e6 2 a 30-5 e ee e.ge e.et 1 m to-I 3 m R0-9 e a ge sum 4 S a 10-7 2.9 3.e 2.e g

g

• I But S 3 e 89-7 2.9 4.0 1.e e

e. 3

0. 3 2 m te"I S et e a 38* 3 S e 30-3

& a 40* 3 6 a te*4 y a te 4

4

9. 3 2 e 10-3 e s 10 e a B6-8 I a Re-3 9 a te"I 7 a R$*S g a Ec pm 4 4 m 14 32.9 16.6 3e e

e/m 4

4

...e--

2..r5 2. irs

.. r5 2..e-S

..e

..0 es.~

,e..

2 10- >

S..~

1..r.

S..O

• i e.e--

O e..e S

.. S

.,. =

e e

....-u e

e

...e4

.. tv.

,e v.

...r>

..r

..,~

S

.. S

...~

2..

..S

..S

,5

,,e

..e

...e

...e

...e e. 05 e.S 10

• I

-I gnut 2

& a 184 3D.9 S.e 2.9 e

3e 4.9

? m 16 e.90 e.Se

6. 30 0.10 e.81 4 s te Em B 2 e Eg-S 3D.9 3.e 2.9 25 20 1.0

? e 18-1 e.no 0.1e

e. no 0.01 0.02 e a 10

-I 8

Esa 4 2 e 194 S.S 3.0 2.9 25 e,a s.e 1 e 10 a a le**

S a 30~ 3 e a no

& a 1e4 6 m 10 3 m to 4

6 e 10'II 4 m to e o 10-12 e, ge-14 e

e 4

2 m to Wim S 1 e 18-8 3.S S.e era 11e sn/a S e 10 Een e hesse a se ts. to.m.,o ere ee ena eene.no e mmas o n..

Le a,,-ass vis.

Eb8 -

ad = to W otth e:

- Seelmee in the celeetettene. any eteer to moglieable ansee its reteeee fseetSea Se setet. testy eme14 See all Berge setemme estegoriet.

Ee9 mestades an, % es, ses. Te.

EG Ematuene T. Ee. ar. % Co. Dr. 34. % p% an. ca.

fel asetAant etthtm Fuh I

_, houe ese A&stimet emergy releemme that effect fab % categmry 0

to enhe1*GM Rees fue la esth a yeekeh&11ty of 4 e le"Iper reagent-year and 20 s 10 ste/hg and pus 19 otth a yeekshality of S a RS-I per remetme'-year and 529 m tee etg/hr.

EfB met myyaam..

09I & Ie N Mb So M Ee ylese of sees Wing the Qat of a potentiel coatsw W. any Bugest en the sonetts usata he etsp and _

Geo.

N 2

en C

W A

to CD s.

I O

O s

to E *3 a

W 9

O O

3

[T h

Om Nw O

NW m

I I

APPENDIX B APPLICA8ILITY OF NLREG-0933 ISSLES TO OPERATING AND FUTURE PLANTS This appendix contains a listing of those residual GSIs that are applicable to operating and future plants and includes: Issues that have been resolved with requirements [ NOTE 3(a), I]; USI, HIGH-and EDIUN-priority issues scheduled for resolution; nearly-resolved issues scheduled for resolution (NOTES 1 and 2); and issues that are scheduled for prioritization (NOTE 4). The priority designations for all issues are consistent with those Itsted in Table i

II of the Introduction. In accordance with 10 CFR 52.47(a)(1)(tv), any future appilcation for <iesign certification must contain proposed technical resolutions for the issues in this ilsting that are designated USI, HIGH, EDILM, NOTE I, and NOTE 2. Also included in this ilsting are those GSIs that were either priorittred or resolved with no ispect on operating plants, but contain monumendations for future plants (NOTE 6).

y Leaend NOTES:

1

- Possible Resolution Identified for Evaluation 2

- Resolution Available (Documumted in NUREG, NRC Memorandum, SER or equivalent)

L 3(a) - Resolution Resulted in the Estabitstunent of New Regulatory Requirements (Rule, Regulatory Guide, SRP Change, or

[

equivalent) 4

- Issue to be Priorittred in the Future r

6

- New Requirements for Future Plants Recomumended

[

B8W

- Babcock & Wilcox Company CE

- Com6ustion Engineering Campany GE

- General Electric Company HIGH

- High Safety Priority I

- Resolved TMI Action Plan Item with Inglementation of Resolution Mandated by NLREG-0737 ED!LM

- Meditse Safety Priority WA

- Multiplant Action NA

- Not Appilcable TBD

- To Be Determined USI

- Unresolved Safety Issue W

- Westinghouse Electric Corporation h

f 2

c

^

X3 o

3 O

• ~

w j

L P

h

.m.

.a --

em

booendix 0 (Continued 1 Ar. tion Operating Future Safety Affected NSS$ Vendor Operating Plants -

Plants-N Plan Iten/

Priority /

Plants-Effective Effective

$ Issue No.

Title Status 8Wt PWR MPA No.

Date Date D,

o TMT ACTION PLAN ITEMS L.A OPERATING PERSONNEL W

Operatina Personnel and Staffina I> 1.1 Shift Technical Advisre I

All All F-01 09/13/79 09/27/79 1.A.I.2 Shif t Supervisor Administrative Duties I

All All 09/13/79 09/27/19 I.A.I.3 shift Manning I

All All F-02 07/31/80 06/26/80 I.A.I.4 Long-Tem Upgrading NOTE 3(a)

All All 04/28/83 04/28/83 W

Trainina and Qualifications of Operatt w Personnel I.A.2.1 Isumediate Upgra. ding of Operator and Senior Operator Training and Qualifications I.A.2.l(l)

Qualificat wis - Experience I

All All F-03 03/28/80 03/28/80 I.A.2.l(2)

Training I

All All F-03 03/28/80 03/28/80

!.A.2.l(3)

Facility Certification of Competence and Fitness of I

All All F-03 03/28/80 03/28/80 Applicants for Operator and Senior Ope?ator Licenses I.A.2.3 Ackninistration of Training Programs I

All All 03/28/80 03/28/80 03 I.A.2.6 Long-Tern Upgrading of Training and Qualifications M I A.2.6(1)

Revise Regulatory Guide 1.8 NOTE 3(a)

All All T80 05/--/87 LA_J Licensina and Reaualification of Operatino Personnel I.A.3.1 Revise Scope of Criteria for i.icensing Examinations I

All All 03/28/80 03/28/80 I.A.4 Simulator t)se and Develoosent I.A 4.1 Initial Simulator Improvement I.A.4.l(2)

Interim Changes in Training Simulators NOTE 3(a)

All A'1 04/--/81 03/28/81 I.A.4.2 Long-Ters Training Simulator Upgrade I.A.4.2(1)

Research on Training Simulators NOTE 3(a)

All All 04/--/87 04/--/87 I.A.4.2(2) tygrade Training Simulator Standards MOTE 3(a)

All All 04/--/81 04/--/81 I.A.4.2(3)

Regulatory Guide on Training Simulators NOTE 3(a)

All All 04/--/81 04/--/81 1.A.4.2'4)

Review Simulators for Conformance to Criteria NOTE 3(e)

All All 03/25/87 03/25/87 LC OP_ERATING PROCEDURES I.C.1 Short-Tem Accident Analysis and Procedures Revision I.C.l(l)

Small 8reak LOCAs I

All All 09/13/19 09/13/79 33 32 I.C.l(2)

Inadequate Core Cooling I

All All F-04 09/13/19 09/13/79 M

I.C.l(3)

Transients and Accidents I

All All F-05 09/13/79 09/27/79 d-1.C.2 Shift and Reitef Turnover Procedures I

All All 09/13/79 09/27/19 8

I.C.3 Shif t Supervisor Responsibilities I

All All 09/13/79 09/27/79 o

O I.C 4 Control Room Access I

All All 09/13/79 09/27/79 3

W I.C.S Procedures for Feedback of Operating Experience to I

All All F-06 0S/07/80 06/26/80

1. 1 ant Staff O

O O

l m

f Assendtr 8 (Continued)

Operating future O Action Safety Affected NSSS Vender.

Operating Plants -

Plants-

% Plan Item /

Priority /

_ Plants-Effective Effective y Issue No.

Title Status INR PWR WA No.

Date Date no cm I.C.6 Procedures for Vertfication of Correct Perfonmence of 1

All All F-07 10/31/80 10/31/80 l

Operating Activities I.C.7 NS;S Vendor Review of Procedures I

All All -

IIA 06/26/80 l.C.8 Pilot leonttoring of Selected Emergency Procedures for I

All All NA 06/26/80 leear-Term Operating Ltcense Appltcants I.C.9 Long-Term Program Plan for t>;rading of Procedures NOTE 3(a)

All All 09/13/19 06/--/85 L.R CONTROL RolBI KSIGN j

I.D.1 Control Room Design Reviews I

All All F-08 06/26/80 06/26/80 I.D 2 Plant Safety Parameter Disclay Console I

All All F-09 06/26/80 06/26/80 I.D.5 Improved Control Room Instrumentation Research I.D.5(2)

Piant Status and Post-Accident Monitoring NOTE 3(a)

All All 11 4 12/--/80 LF_

OWLITY ASS WARCL I.F.2 Develop More Detailed QA Criteria I F.2(2)

Include QA Personnel in Review and Approval of Plant NOTE 3(a)

All All NA 07/--/81 Procedieres I.F.2(3)

Include QA Personnel in All Design, Construction, NOTE 3(a)

All All NA 07/--/81 y

Installation. Testing, and Operation Activittes I.F.2(6)

Increase the Size of Licensees

• QA Staff NOTE 3(a)

All All IIA 07/ - /81 I.F.2(9)

Clarify Organizational Reporting Levels for the QA NOTE 3(a)

All All NA 07/--/81 Organization L.i PREOPERATIONAL AND LOW-POWER TESTING I.6.1 Training Requirements I

All All NA 06/26/80 I.6.2 Scope of Test Program NOTE 3(a)

All All NA 07/--/81 ILE CONSIKRATION OF DEGRADED OR IELTED CORES IN SAFETY REVIEW L

II.B.1 Reactor Coolant System Vents I

All All F-10 09/13/79 09/27/79 11.8.2 Plant Shielding to Provide Access to Vital Areas and I

All All' F-Il 09/13/19 09/27/79 Protect Safety Equipment for Post-Accident Operation 11.8.3 Post-Accident SampIfng I

All All F-12 09/13/79 09/27/19

!!.8.4 Training for Nitigating Core Damage I

All All F-13 03/28/80 03/28/80 II.B.6 Risk Reduction for Operating Reactors at Sites with NOTE 3(a)

All All T80

.NA M

y High Population Densities k

m II.B.8 Rulemaking Proceeding on Degraded Core Accidents h0TE 3ta)

All All T80 01/25/85 i

S 1.L.R REACTOR COOLANT SYSTEN RELIEF AND SAFETY VALVES j

$ 1I.D.1 Testing Requirements I

All All F-14 09/13/79 09/27/79 y

f W II.D.3 Relief end hfety Valve Position Indication I

All All 07/21/19 09/27/79 I

l t

i

I l

{

Anoendix 9 (Continuadt Operating Future

$ Action Safety Affected NSSS Vendor Operating Plants -

Plants-N Plan item /

Priority /

Plants-Effective Effective

$ Issue No.

Title Status BWR PWR MPA No.

Date Date N

10 m

J,1,1 SYSTEM DESIGN II.E 1 Auxiliary Feedrater System

!!.E.1.1 Auxiliary Feedwater System Evaluation I

MA All FIS 03/10/80 03/10/80 II.E.1.2 f.uxiliary Feedwater System Automatic Initiation and I

NA All F-16, F-17 C1/13/19 09/27/79 Flow Indication II.E.1.3 Update Standard Review Plan and Develop Regulatory NOTE 3(a)

All All NA 07/~/81 Guide II.E,3 Decay Heat Removal II.E.3.1 Reliabiltty of Power Sur71ies for Natural Circulation I

NA All 09/13/19 09/27/79 II,E.4 Containment Desian 11.[.4.1 Dedicated Penetrations 1

M1 All F-18 09/13/79 09/27/79 II.E.4.2 Isolation Dependability I

All All F-19 09/13/79 09/27/79 II.E.4.4 Purging II.E.4.4(1)

Issue Letter to Licensees Requesting Limited Purging NOTE 3(a)

All All 11/28/78 NA

?

II.E.4.4(2)

Issue letter to Licensees Requesting Information on NOTE 3(a)

All All 10/22/79 NA tn Isolation Letter i

II.E.4.4(3)

Issue tetter to Licensees on valve Operability NOTE 3(a)

All All 09/27/79 NA II.E.5 Destan Sensitivity of B&W Reactors II.E.5.1 Design Evaluation NOTE 3(a)

NA B&W II.E.5.2 B&W Reactor Transient Response Task Force NOTE 3(a)

MA B&W II E.6 In Situ Testina of Valves II.E.6.1 Test Adequacy Study NOTE 3(a)

All All 06/--/89 06/--/89 JL.F.

INSTRUMENTATION AND CONTROLS II.F.I Additional Accident Monitoring Instrumentation 1

All All F-20. F-21 09/13/79 09/27/79 F-22, F-23 F-24, F-25 II.F.2 Identif tcation of and Recovery from Conditions 1

All All F-26 070/2/79 09/27/79 Leading to Inadequate Core Cooling II.F.3 Instruments for Menitoring Accident Conditions NOTE 3(a)

All All NA 12/--/80 It

$ jl s ELECTRICAL POWER Q

a w

11.G.1 Power Supplies for Pressurizer Relief Valves, Block i

NA All 09/13/79 09/27/79 i

s Valves, and Level Indicators o

O 3

e 44 (c) to O

O O

Ascendix 8 (Continued)

Operating future

$ Action Safety Affected NSSS Vendor Operating Plar.ts -

Plants-N Plan Item /

Priority /

Plants-Effective Effective

$ Issue No.

Title Status BWR fnm W A No.

Date Date D

m M

GENERAL IDFLICATIONS OF TMI FOR DESIGN A M CONSTRUCTION ACTIVITIES II,J.4 Revise Deficiency Resortine Reautrements i

II.J.4.1 Revise Deficiency Reporting Requirements NOTE 3(a)

A11 All 07/31/91 07/31/91 11 K KASIRES TO NITIGATE SMALL-BMAK LOSS-OF-COOLANT ACCIDENi$ A W LOSS-OF-FEEDWATER ACCIDENTS II.K.1 IE Bulletins II.K.l(!)

Review TMI-2 PNs and Detalt=J N 1ogy of the NOTE 3(a)

All All 03/31/80 NA TNI-2 Accident II.K.l(2)

Review Transients Stellar to TMI-2 inat Have NOTE 3(a)

NA 84W 03/31/80 NA Occurred at Other Facilities and NRC Evaluation of Davis-8 esse Event II.K.l(3)

Review Operating Procedures for Recognizing, NOTE 3(a)

NA Kil 03/31/80 NA Preventing, and Mitigating Vold Forumtion in Transients and Accidents

?

II.K.1(4)

Review Operating Procedures and Training NOTE 3(a)

All All 03/31/80 NA CD Instructions E

II.K.l(S)

Safety-Related valve Position Description NOTE 3(a)

All All 03/31/80 03/31/60

)

II.K.l(6)

Review Contalrument Isolation Initiation Design NOTE 3(a)

All All 03/31/80 NA r

and Procedures II.K.l(7)

Implement Positive Position Controls on Valves NOTE 3(a)

NA BaW 03/31/80 NA That Could Compromise or Defeat AFW Flow f

II.K.l(8) leplement Procedures That Assure Two I.44.t NOTE 3(a)

NA 88W 03/31/80 NA 100% AFW Flow Paths i

II.K.I(9)

Review Procedures to Assure That Radioactive NOTE 3(a)

All All 03/31/80 NA Liquids and Gases Are Not Transferred out of Contatrument Inadvertently II.K.1(10)

Review and Nodify Procedures for Removing Safety-NOTE 3(a)

All All 03/31/80 03/21/80 Related Systems from Service II.K.l(11)

Make All Operating and Naintenance Personnel NOTE 3(a)

All All 03/31/80 NA 3

Auare of the Seriousness and Consequences of the j

Erroneous Actions Leading up to, and in Early E

Phases of, the TNI-2 Accident II.K.1(12)

One Hour Notification Requirement and Continuous NOTE 3(a)

All All NA Communications Channels II.K.l(13)

Propose Technical Specification Changes Reflecting NOTE 3(a)

All All 01/01/81 01/01/81 x

Implementation of All Bulletin Items g

II.K.1(14)

Review Operating Modes and Procedures to Deal with NOTE 3(a)

GE CE, W 03/31/80

1. A y

o Significant Amounts of Hydrogen e

II.K.1(IS)

For Fact 11 ties with Non-Automatic AFW Initiation, NOTE 3(a)

NA CE. ~W NA o"

O Provide Oedicated Operator in Continuous y

Communication with CR to Operate AFW g

9 i

6

Anoendtx 8 fContinued)

Operating Ftture

$ Action Safety Affected NSSS Vendor Operating Plants -

Plants-N Plan Item /

Priority /

Plants-Effective Effective Issue No.

Title Status BWt PWR W A No.

Date Date N

!!.K.l(16)

Iglement Procedures That Identify PRZ PORY "Open" NOTE 3(a)

NA CE. W NA Indications and That Direct Operator to Close knually at " Reset" Setpoint II.K.1(17)

Trip PZR Level Sistable so That PZR Low Pressure NOTE 3(a)

NA V

Vill Initiate Safety Injection II.K.l(18)

Develop Procedures and Train Operators on Methods NOTE 3(a)

NA BW NA of Estabitshing and h intaining Natural Circulation II.K.1(19)

Describe Design and Procedure Modifications to NOTE 3(a)

NA BW 03/31/80 NA Reduce Likelihood of Automatic PZR PORT Actuation in Transients ll.K.l(20)

Provide Procedures and Training to Operators for NOTE 3(a)

NA BW 03/31/80 03/31/80 Prong,L Manual Reactor Trip for LOFW, TT. MSIV Closure, LOOP, LOSG Level, and LO PIR Level II.K.I(21)

Provide Automatic Safety-Grade Anticipatory Reactor NOTE 3(a)

NA BW 03/31/83 03/31/80 1 rip for LOFV, TT. cr Significant Decrease in SG Level II.K.l(22)

Describe Automatic and Manual Actions for Proper NOTE 3(a)

All NA 03/31/80 03/31/80 Functioning of Aux 11tary Heat Renoval Systems When y

FV System Not Operable cn II.K.l(23)

Describe Uses and Types of RV Level Indication for NGTE 3(a)

All NA 03/31/80 03/31/80 Automatic and Manual Initiation Safety Systems II.K.l(24)

Perform LOCA Analyses for a Range of Small-Break NOTE 3(a)

NA All NA Sizes and a Range of Time Lapses Between Reactor Trip and RCP Trip II.K.l(25)

Develop Operator Action Guioellnes NOTE 3(a)

NA All NA II.K.l(26)

Revise Emergency Procedures and Train R0s and SR0s NOTE 3(a)

NA All NA II.K.l(27)

Provide Analyses and Develop Guidelines and NOTE 3(a)

NA All NA Procedures for inadequate Core Cooling Conditions ll K.l(28)

Provide Design That Will Assure Automatic RCP Trip NOTE 3(a)

NA All 01/01/81 01/01/82 for All Circumstances Where Required II.K.2 Commission Orders on BW Plants II.K.2(1) tipgrade Timeliness and Reliability of AFV System NOTE 3(a)

NA BW NA II.K.2(21 Procedures and Training to Initiate and Control NOTE 3(a)

NA BW NA AFW Independent of Integrated Control System II.K.;(3)

Hard-Wired Control-Grade Anticipatory Reactor Trips NOTE 3(a)

NA BW NA II.K.2(4)

Small-Break LOCA Analysis, Procedures and Operator NOTE 3(a)

NA BW NA Training II.K.2(S)

Caglete TMI-2 Simulator Training for All Operators NOTE 3(a)

NA BW NA y

2c-II.K.2(6)

Reevaluate Analysis for Dual-Level Setpoint Control NOTE 3(a)

NA BW NA y

II.K.2(7)

Reevaluate Transient of September 24, 1977 NOTE 3(a)

NA BW NA 7

o II.K.2(9)

Analysis and Upgrading of Integrated Control System I

NA BW F-27 01/01/81 01/01/81 y

II.K.2(10)

Hard-Wired Safety-Grade Anticipatory Reactor Trips I

NA BW F-28 01/01/81 01/01/81 Q

e II.K.2(II)

Operator Training and Drilling I

NA BW F-29 01/01/81 01/01/81 y

II.K.2(13)

Thermal-Mechanical Report on Ef fect of HPI on Vessel I

NA BW F-30 01/01/81 01/01/81 y

Integrity for Small-Break LOCA With No AFW 9

O O

Aooendix B (Continued)

Operating Future

@ Action Safety Affected RSSS Vendor Operating Plants -

Plants-N Plan Item /

Priority /

Plants-Effective Effective

$ Issue No.

Ittle Status BWR PWR MPA No.

Date Date D.

a II.K.2(14)

Dumonstrate That Predicted Lift Frequency of PORVs I

NA B&W F-31 01/01/81 01/01/81 and SVs Is Acceptable II.K.2(15)

Analysts of Effects of Slug Flow on Once-Through I

HA B&W 06/01/80 06/01/80 Steam Generator Tees After Prisery System Votding II.K.2(16)

Impact of RCP Seal Damage Following Small-Break I

NA BRM F-32 06/01/80 06/01/80 LOCA With Loss of Offsite Power II.K.2(17)

Analysts of Potential Voiding in RCS During I

MA Bau F-33 NA Anticipated Transtents II.K.2(19)

Benctunerk Analysts of Sequential ATV Flow to Once-I N4 B&W F-34 01/01/81 NA Through Steam Generator II.K.2(20)

Analysis of Steam Response to Small-Break LOCA I

NA B&W F-35 01/01/81 NA That Causes System Pressure to Exceed PORV Setpoint II.K.2(21)

LOFT L3-1 Predictions NOTE 3(a)

NA B&W NA II.K.3 Final Recommendations of Bulletins and Orders Task Force II.K.3(1)

Install Automatic PORV Isolation System and Perform I

NA All F-36 07/01/81 07/01/81 Operational Test II.K.3(2)

Report on Overall Safety Effect of PORV Isolation I

NA All F-37 01/01/81 01/01/8'

?

System U3 II.K.3(3)

Report Safety and Relief Valve Failures Pragtly I

All All F-38 04/01/80 04/01/80 and Challenges Annually II.K.3($)

Automatic Trip of Reactor Coolant Pu g s I

NA All F-39, 6-01 01/01/81 01/01/81 II.K.3(7)

Evaluation of PORV Opening Probability During I

NA B&W 01/01/81 01/01/81 Overpressure Transient II.K.3(9)

Proportional Integral Derivative Controller I

NA V

F-40 07/01/80 07/01/80 Modification II.K.3(10)

Anticipatory Trip Modification Proposed by Some I

NA W

F-41 Licensees to Confine Range of Use to High Power Levels II.K.3(ll)

Control Use of PORV Suppiled by Control Cogonents.

I All All Inc. Untti Further Review Co glete II.K.3(12)

Confirm Ertstence of Anticipatory Trip Upon Turbine I

NA W

F-42 07/01/90 07/01/80 Trip II.K.3(13)

Separation of HPCI and RCIC System Initiation Levels I

GE NA F-43 10/01/80 10/01/80 II.K.3(14)

Isolation of Isolation Condensers on High Radiation I

GE NA F-44 01/01/81 NA II.K.3(15)

Modify Break Detection Logic to Prevent Spurious I

GE NA F-45 01/01/81 01/01/81 Isolation of HPCI and RCIC Systems II.K.3(16)

Reduction of Challenges and Failures of Relief I

GE NA F-46 01/01/81 01/01/81 Valves - Feasiblitty Study and System Modification

Z3 II.K.1(17)

Report on Outage of ECC Systems - Licensee Report I

GE NA F-47 01/01/81 01/01/81

Z3 and Technical Spectitcation Changes II.K.3(18)

Modification of ADS Logic - Feasibility Study and I

GE NA F-48 01/01/81 01/01/81 f.,

e Modification for Increased Olversity for Some o"

Event Sequences w

II.K.3(19)

Interlock on Rectreulation Pump Loops I

GE NA F-49 01/01/81 NA e-a N

II.K.3(20)

Loss of Service Water for Big Rock Point I

GE NA 01/01/81 NA

Anoendix B fContinued)

Operating Future

$ Action Safety Affected NSSS Vendor Operating Plants -

Plants-N Plan item /

Priority /

Plants-Effective Effective

$ Issue No.

ittie Status 3WR PWR MPA No.

Date Data D

a II.K.3(21)

Restart of Core Spray and LPCI Systems on Low I

E NA F-50 01/01/81 01/01/81 Level - Design and Modtfication II.K.3(22)

Automatic Switchover of RCIC Syst=m Suction -

I E

NA F-51 01/01/81 01/01/81 Verify Procedures and Modify Design II.K.3(24)

Confira Adequacy of Space Cooling for HPCI and I

E N4 F-52 01/01/82 01/01/82 RCIC Systems II.K.3(25)

Effect of Loss of AC Power on Pimp Seals 1

E NA F-53 01/01/82 01/01/82 II.K.3(27)

Provide r m Reference Level for Vessel Level I

E NA F-54 10/01/80 10/01/80 Instrumentation II.K.3(28)

Study and Verify Qualification of Accumulators I

E NA F-55 01/01/82 01/01/82 on ADS Valves II.K.3(29)

Study to Demonstrate Performance of Isolation I

E NA F-56 04/01/81 NA Condensers with Non-Condensibles II.K.3(30)

Revised Small-8reak LOCA Methods to Show Comoliance I

All All F-57 01/01/83 01/01/83 with 10 CFR 50. Appendix K II.K.3(31)

Plant-Specific Calculations to Show Cor.pilance with All All F-58 01/01/83 01/01/83 10 CFR 50.46 II.K.3(44)

Evaluation of Anticipated Transients with Single I

GE NA F-59 01/01/81 01/01/81

?

Failure to Verify No Significant Fuel Failure cn II.K.3(45)

Evaluate Depressuritation with Other Than Full ADS I

E NA F-60 01/01/81 01/01/81 II.K.3(46)

Respor:se to List of Concerns from ACRS Consultant I

E RA F-61 07/01/80 07/01/80 II.K.3(57)

Identify Water Sources Prior to Manual Activation I

E NA F-62 10/01/80 NA of ADS III.A EMfRGENCY PREPAREDNESS AND RADIATION EFFECTS III.A.!

Imorove Licensee Eneroency Preparedness - Short Tere III.A.1.1 Upgrade Emergency Preparedness III.A.I.1(1)

Implement Action Plan Requirements for Promptly I

All All 10/10/79 08/19/80 Isproving Licensee Emergency Preparedness III.A.I.2 Upgrade Licensee Emergency Support Facilities III.A.I.2(1)

Technical Support Centar 1

All All F-63 09/13/79 09/27/79 Ill.A.1.2(2)

On-Site Operational Support Center I

All All F-64 09/13/79 09/27/79 III.A.I.2(3)

Near-Site Emergency Operations Facility I

All All F-65 09/13/19 09/27/79 III.A.2 Improvino Licensee Emeroency Preparedness-Lono Term III.A.2.1 Amend to CFR 50 and to CFR 50, Appendtx E III.A.2.l(l)

Publish Proposed Amendnents to the Rules NOTE 3(a)

All All III.A.2.l(4)

Revise Inspection Program to Cover Upgraded I

All All F-67

o Requirements y

III.A.2.2 Development of Guidance and Criteria I

All All F-68 cn e

III.A.3 Inerovino NRC Emeroency Preparedness o

8 III.A.3.3 Ccaununications g

III.A.3.3(1)

Install Direct Dedicated Telephone Lines NOTE.S(a)

All All O

O O

/~

f%

m k

/

V Accendix 5 iConttnued)

Operating Future gm Action Safety Affected NSSS Vendor Operating Plants -

Plasts-N Plar Item /

Priority /

Plants-Effective Effective

$ Issue flo.

Title Status BWt PWt frA No.

Date Date N$

III.A 3.3(2)

Obtain Dedicated, Short-Range Radio r e tcation NOTE 3(a)

All All Systems llL.D RADIATION PROTECTION III.D.I Radiatton Source Control 111.D.1.1 Primary Coolant Sources Outside the Containment Structure III.D.I.l(1)

Revleu Infomation Submitted by Licensees Pertaining I

All All 07/02/79 09/27/19 to Reducing Leakage from Operating Systems III.D.3 Worker Radiation Protection fi m

..t 111.D.3.3 Inplant Radiation Monitoring III.D.3.3(1)

Issue Letter Requiring Igroved Radiation Sampling I

All All F-69 09/13/19 09/27/79 Instruentation III.D.3.3(2)

Set Criteria Requiring Licecsees to Evaluate Need for NOTE 3(a)

All All 09/13/79 09/27/19 Additional Survey Equipment III.D.3.3(3)

Issue a Rule Change Providing Acceptable Methods for NOTE 3(a)

All All 09/13/19 09/27/19 Calibration of Radiation-Monttoring Instruments CD III.D.3.3(4)

Issue a Regulatory Guide NOTE 3(a)

All All 09/13/79 09/27/79 III.D.3.4 Control Room Habitability I

All All F-70 05/07/80 06/26/80 TASK ACTION PLAN ITEMS A-1 Water Hammer (fomer USI)

NOTE 3(a)

All All NA 03/15/84 A-2 Asymmetric Bloudoun Loads on Reactor Primary Coolant NOTE 3(a)

NA All D-10 01/--/81 01/--/81 Systems (fomer USI)

A-3 Westinghouse Steam Generator Tube Integrity (fonner USI)

NOTE 3(a)

NA y

04/17/85 04/17/85 A-4 CE Steam Generator Tube Integrity (fonner USI)

NOTE 3(a)

NA CE 04/17/85 04/17/85 A-S 88M Steam Generator Tube Integrity (fomer USI)

IIOTE 3(a)

NA 88M 04/17/85 04/17/85 A-6 Mark I Short-Tem Program (fomer USI)

NOTE 3(a)

GE NA 12/--/77 NA A-7 Mark I long-Tem Program (fomer USI)

NOTE 3(a)

GE IIA D-01 08/--/82 08/--/82 A-8 Park II Contatrument Pool Dyannic Loads - Long Tene NOTE 3(a)

GE NA 08/ - /81 08/--/81 Program (former USI)

A-9 AfwS (fonner USI)

IIOTE 3(a)

All All 06/26/84 06/26/84 A-10 BWt Feedsater llozzle Cracking (fonner USI)

NOTE 3(a)

All NA 8-25 11/--/80 11/--/80 A-Il Reactor Vessel Materials Toughness (fonner USI)

NOTE 3(a)

All All 10/--/82 NA A-12 Fracture Toughness of Steam Generator and Reactor NOTE 3(a)

NA All NA TBD Coolant Pump Supports (fomer USI)

{

c-A-13 Snubber Operability Assurance NOTE 3(a)

All All 1980 1980 A-16 Steam Effects on 8Wt Core Spray Distribution NOTE 3(a)

GE NA D-12 NA A-24 Qualtftcation of Class IE Safety Related Equipment NOTE 3(a)

All All B-60 08/--/81 08/--/81 E

O e

(former USI)

A-25 Non-Safety Loads on Class IE Power Sources IIOTE 3(a)

All All 09/--/78 y

A-26 Reactor Vessel Pressure Transient Protection NOTE 3(a)

NA All 8-04 09/--/78 09/--/78 y

(fonner USI)

a Acoendix B (Continued) g Operating Future m Action Safety Affected NSSS Vendor Operating Plants -

Plants-N Plan Item /

Priority /

Plants-Effective Effective

$ Issue No.

Title Status BWR PWR MPA No.

Date Date N

40 01 A-28 Increase in Spent Fuel Pool Storage Capacity NOTE 3(a)

All All 04/11/78 NA A-31 Rm Shutdown Requirements (fomer USI)

NOTE 3(a)

All All 05/--/78 10/01/78 A-35 Adequacy of Offsite Power Systems NOTE 3(a)

All All 06/02/77 1980 A-36 Control of Heavy Loads hear Spent Fuel (former USI)

NOTE 3(a)

All All C-10, C-15 07/--/80 07/--/80 A-39 Determination of Safety Reitef Valve Pool Dynamic NOTE 3(a)

GE RA 02/29/80 09/30/80 Loads and Tegerature Limits (former USI)

A-40 Seismic Design Criteria (former USI)

NOTE 3(a)

All All TBD 09/--/89 A-42 Pipe Cracks in Bolling Vater Reactors (former USI)

NOTE 3(a)

All NA 8-0$

02/--/81 02/--/81 A-43 Contatrument Emergency Sug Perfomance (former US1)

NOTE 3(a)

NA All NA 11/-/85 A-44 Station Blackout (former USI)

NOTE 3(a)

All All TBD 06/--/88 A-46 Seismic Qualification of Eoulpment in Operating Plants NOTE 3(a)

All All 02/--/87 NA (former USI)

A-47 Safety Igitcations of Control Systems (former USI)

NOTE 3(a)

All All 09/20/89 09/20/89 A-48 Hydrogen Control Measures and Effects of Hydrogen Burns NOTE 3(a)

All V

12/--/81 12/--/81 on Safety Equipment A-49 Pressurtred Thermal Shock (former USI)

NOTE 3(a)

NA All A-21 TBD 07/--/85 8-10 Behavior of BWR Mark III containments NOTE 3(a)

GE NA NA 09/--/84 B-11 Criterta for Safety-Related Operator Actions MEDIUM All All TBD TBD y

B-36 Develop Design. Testing, and Maintenance Criteria for NOTE 3(a)

All All 03/--/78 a

Atmosphere Cleanup System Air Flitration and Adsorption g

Units for Engineered Safety Feature Systems and for Normal Ventilation Systems B-SS I m roved Reliability of Target Rock Safety Relief MEDIlm All NA TBD TBD Valves B-$6 Slesel Reliability NOTE 3(a)

All All D-19 06/--/93 06/--/93 B-61 Allowable ECCS Equipment Outage Periods MEDItM All All TBD TBD B-63 Isolation of Low Pressure Systems Connected to the NOTE 3(a)

All All 04/20/81 Reactor Coolant Pressure Boundary B-64 Decommissioning of Reactors EDTE 3(a)

All All 06/27/88 NA B-66 Control Room Infiltration 84easurements NOTE 3(a)

All All NA 07/--/81 C-1 Assurance of Continuous Long Term Capablitty of NOTE 3(a)

All All 05/27/80 05/27/80 Hermetic Seair on Instrumentation and Electrical Ecutoment C-10 Effective Operation of Contairment Sprays in a LOCA NOTE 3(a)

All All NA C-17 Interim Acceptance Criteria for Solidification Agents NOTE 3(a)

All All 12/27/82 12/27/82 for Radioactive Solid Wastes NEW GENERIC ISSUES

x2 h

23.

Reactor Coolant Pump Seal Failures HIGH MA an TBD TBD p

25.

Automatic Air Header Dump on BWR Scram System NOTE 3(a)

All NA 01/09/81 01/09/81 y

cn 9

O O

40.

Safety Concerns Associated with Pipe Breaks in the BWR NOTE 3(a)

All NA B-65 08/31/81 08/31/81 W

Scram System

~"

41.

BWR Scram Discharge Volune Systems NOTE 3(a)

All NA B-58 12/09/80 NA O

O O

m._

m.._

I Anoendix 9 (Continued)

Operattng Future Q

Action Safety Affected NSSS Vender Operating Plants -

Plants-Plan Item /

Priority /

Plants-Effective Effective Issue No.

Title Status Blst PlR IFA No.

Date Date 43.

Reliability of Air Systems NOTE 3(a)

All All 08/08/88 08/08/88 45.

Inoperability of Instriamentation Due to Extreme Cold NOTE 3(a)

All All NA 09/01/83 Weather 51.

Proposed Requirements for leproving the Reliability of NOTE 3(a)

All All 07/18/89 07/18/89 Open Cycle Service Water Systems 67.

Steam Generator Staff Actions 67.3.3 leproved Accident Monitoring NOTE 3(a)

All All A-17 12/11/82 12/17/82 70.

PORV and Block Valve Reliability NOTE 3(a)

NA All 06/25/90 06/25/90 73.

Detached Thermal Sleeves NOTE 3(a)

NA V

NA 75.

Generic laplications of ATWS Events at the Saleum NOTE 3(a)

All All 8-76, 8-77, 07/08/83 TBD Nuclear Plant 8-78, 8-79, S-80, 8-81, 8-82, 8-85, 8-86, B-87, 8-88, 8-89, l

8-90, 8-91, 8-92, 8-93 78.

Monitoring of Fatigue Transient Limits for Reactor teEDILM All All T80 T80 Coolant System y

86.

Long Range Plan for Dealing with Stress Corrosion NOTE 3(a)

All NA 8-84 TBD T80 Cracking in 8tst Piping 87.

Failure of HPCI Steam Line Without Isolation NOTE 3(a)

All All 06/28/89 06/28/89 89.

Stiff Pipe Clamps NOTE 6 All All NA NA T80 93.

Steam Binding of Auxiliary Feedmater Piamps NOTE 3(a)

NA All 10/--/85 10/ - /85 r

94 Additional Low Temperature Overpressure Protection NOTE 3(a)

NA CE, W 06/25/90 06/25/90 for Light Water Reactors 99.

RCS/RMt Suction Line Valve Interlock on Plats NOTE 3(a)

NA All 10/11/88 10/11/88 103.

Design for Probable Marianan Precipitation NOTE 3(a)

All All 10/19/89 10/19/89

'l 118.

Tendon Anchorage Failure NOTE 3(a)

All All NA NA 07/--/90 124.

Auxiliary Feedmater Systein Reliability NOTE 3(a)

All All T80 T80 i

128.

Electrical Power Reliability NOTE 3(a)

All All 04/29/91 04/29/91 130.

Essential Service Water Pump Failures at Multiplant NOTE 3(a)

NA All 09/19/91 09/19/91

[

Sites 155.

Generic Concerns Arisino fror TMI-2 Cleanuo

[

155.1 More Realistic source Tern Assumptions NOTE 3(a)

All All NA NA 02/--/95 156.

Systematic Evaluation Program 156.6.1 Pipe Break Effects on Systems and Co w.2-+=

NOTE 4 All All T80 T80 j

158.

Performance of Pomer-Operated Valves Under MEDitM All

. All T80 TBD i

Design Basis Conditions

D 2

163.

Multiple Steam Generator Tube Leakage NOTE 4 NA All T80 T80 i,

o 165.

Safety avi Safety / Relief Valve Reliability HIGH All All T80 T80 i

g 166.

Adequacy of Fatigue Life of Metal Components NOTE I All All TBD T80 a

168.

Environmental 1)ualification of Electrical Equipment NOTE I All All T80 T80 o

8 o

l

Accendix B fConti W 1 Operating Fsture

@ Action Safety Affected 11555 Vendor Operating Plants -

Plants-N Plan Item /

Priority /

Plants-Effective Effective

$ Issue No.

Title Status IMt PWR MPA No.

Date Date D,

o 169.

8W HSIV e m Mode Failure Due to Loss of NOTE 4 All NA TBD Tr i Accumulator Pressure 170.

Fuel Damage Criteria for High Burnup Fuel NOTE 2 All All TBD TBD 171.

ESF Failure frau LOOP Subsequent to A LOCA HIGH All All TBD TBD 172.

hittple Systen Responses Program NOTE 2 All All T80 TBD 173.

Saant Fuel Storace Pool 173.A operating Fact 11ttes NOTE 2 All All 173.8 Permanently Shutdoun Factittles NOTE 2 All All 177.

Vehicle Intrusion at TMI NOTE 3(a)

All All HiM UI FACT (Nt$ ISSUES F1 STAFFING AND QUEIFICATIONS 95.1.1 Shift Staffing NOTE 3(a)

All All 01/--/84 01/--/84

?

L i'=

O E

5.

G i

b E

u

=

0 0

0

.... -.. - - -. ~. - _. - -... -.... -.. - -.. - - - -.. - - _ -... ~. ~ -. - -.

1 I

l Revision 1 i

i 4

APPENDIX C i

PRIORITY RANKING NUMERILAL THRESH 0LDS USED IN PRIORITIZATIONS C0HPLETED BEFORE JUNE 30. 1993 I

i i,

I t

}

i i

N 06/30/96 A.C-1 NUREG-0933

l 8

Legend:

'm o

H =HIGH priority sg D

L M

H H

M = MEDIUM priority l

l L = LOW priority i

D = DROP 3

3,000 s

E l

I e

E i

6 3

D L

M M

H

[

i

?

E m

o t

o m

100 t

o ti

>4 f

h 3

E a

D L

L M

H E

E o

10 i

l D

D L

M

~

H i

i 10' 10 19 Man-Rom / Reactor 2

5x10' 5x10 5x10s 5x10' Man-Rem (Total, All Reactors) l 10 10-7 10*

10' Core-Melt /RY

(

8 E

5x1(T' 5x10-*

5x10' 5x10 Core-Melt /Yr C

d Change in Risk

~

i O

O O

\\

. _ _ ~ _ _ _

)

I l

Revision 1 i

I Value/ Impact Score Thresholds To the extent consistent with the safety-importance screening criteria just discussed, the value-impact priority score, S, is translated into priority rankings in accordance with the follow-ing thresholds:

1 l

(a) HIGH priority - At least 3,000 man-rem /$million, 1

(b) MEDIUM priority - Less than 3,000 but at least 100 man-rem /$million (c) LOW priority - Less than 100 but.at least 10 man-rem /

l

$million, (d) OROP category - Less than 10 man-rem /$million.

j a

I l

i I

i 06/30/96 A.C-3 NUREG-0933

Revision 1 O

TABLE I RISK THRESHOLDS (a) The priority rank is always HIGH when any of the following risk (or risk-related) thresholds are estimated to be exceeded (or when extraordinary uncertainty suggests that they may well be exceeded):

(1) 1,000 man rem estimated public dose per remaining reactor lifetime (2) 50,000 man-rem total estimated for all affected reactors for their remaining lifetime (e.g., 500 man-rem / reactor for 100 reactors)

(3) 10 5/ reactor year large-scale core melt (4) 5 x 10 4/ year large-scale core melt (total for all affected reactors)

(b) Always at least MEDIUM priority:

10 or more percent of the always-HIGH criteria (c) Always at least LOW priority:

1 or more percent of the always-HIGH criteria (d) Never higher than MEDIUM priority:

Less than 10% of the always-HIGH criteria (e) Never higher than LOW priority:

Less than 1% of th? always-HIGH criteria (f) Always DROP category:

Less than 0.1% of the always-HIGH criteria 0

06/30/96 A.C-4 NUREG-0933

l O

APPENDIX D RELATED GENERIC ACTIVITIES This appendix documents those activities related to generic issues, i.e., related generic activities (RGA), that did not meet the criteria for designation as generic issues (GI), but were important enough to require the development of Action Plans by NRR to address the concerns. The plan for documenting these RGAs was delineated in SECY-96-107."'

O 1

i O

06/30/96 A.D.0-1 NUREG-0933

C E 001: BOILING WATER REACTOR __JNTERNALS DEERIPTION Significant cracking of the core shroud was first observed at Brunswick-1 in September 1993. The NRC notified licensees of Brunswick's discovery of significant circumferential cracking of the core shroud welds. In 1994, core shroud cracking continued to be the most significant of reported internals cracking. In July 1994, the NRC issued GL 94-03"" which required licensees to inspect their shrouds and provide an analysis justifying continued operation until inspections could be completed.

A special industry review group (Boiling Water Reactor Vessels and Internals Project - BWRVIP) was formed to focus on resolution of reactor vessel and internals degradation. This group was instrumental in facilitating licensee responses to GL 94-03."" The NRC evaluated the BWRVIP reports, submitted in 1994 and early 1995, and all plant responses. All of the plants evaluated were able to demonstrate continued safe operation until inspection or repair on the basis of: (1) no 360* through-wall cracking observed to date; (2) low frequency of pipe breaks; and (3) short period of operation (2 to 6 months) before all of the highly susceptible plants complete repairs of, or inspections to, their core shrouds.

p)

In late 1994, extensive cracking was discovered in the top guide and core plate gv rings of a foreign reactor which was similar in design to GE reactors in the U.S.; however, there have been no observations of such cracking in U.S. plants.

GE concluded that it was reasonable to expect that the ring cracking could occur in GE BWRs that had been in operation for more than 13 years. In the BWRVIP report that was issued in January 1995, ring cracking was evaluated. The NRC concluded that the BWRVIP assessment was acceptable and that top guide ring and core plate ring cracking was not a short-term safety issue. This activity was identified in an NRR memorandum"" to RES in February 1996.

I Many components inside BWR vessels, i.e., internals, are made of materials such as stainless steel and various alloys that are susceptible to corrosion and cracking. This degradation can be accelerated by stresses from temperature and pressure changes, chemical interactions, irradiation, and other corrosive environments.

The action plan"" developed by NRR is intended to encompass the evaluation and resolution of issues associated with IGSCC in BWR internals, including plant-specific reviews and the assessment of the generic criteria proposed by the BWR i

Owners' Group. The staff will continue to assess the scopes that have jet to be submitted by licensees concerning inspections or re-inspections of their core shrouds. The staff will also continue to assess core shroud inspection results and any appropriate core shroud repair designs on a case-by-case basis. The staff will issue separate safety evaluations regarding the acceptability of core shroud inspection results and core shroud repair designs. The staff has been interacting O

with the BWRVIP and individual licensees. In an effort to lower the number of

('j industry and staff resources that will be needed in the future, it is important for the staff to continue interacting with the industry on a generic basis in 06/30/96 A.D.001-1 NUREG-0933

order to encourage them to continue their proactive efforts to resolve IGSCC of BWR internals. The BWRVIP has submitted four generic documents, supporting plant-specific submittals, for staff review. The staff is ensuring that the generic reviews are incorporating recent operating experience on all BWR internals.

CONCLUSION 4

Based on licensee responses to GL 94-03,"" the staff concluded in all cases that licensees have provided sufficient evidence to support continued operation of their BWR units to the refueling outages in which shroud inspections or repairs have been scheduled. In addition, by the end of 1995, industry's special review group that is aggressively pursuing this issue was expected to issue a comprehensive plan addressing cracking in all BWR internals, discussing cracking susceptibility, safety consequences, inspection scope and methodology, fl aw evaluation, repair strategies, and mitigation of degradation.

Almost all BWRs will have completed inspections or repairs of core shrouds during refueling outages by the fall of 1995. Various repair methods have been used to provide alternate load-carrying capability, including preemptive repairs, installation of a series of clamps and use of a series of tie-rod assemblies.

The NRC has reviewed and approved all shroud modification proposals that have been submitted by BWR licensees. Review by NRC continues on individual inspection results and plant-specific assessments.

The industry special review group was expected to issue a comprehensive plan addressing cracking in all BWR internals, discussing cracking susceptibility, safety consequences, inspection scope and methodology, flaw evaluation, repair strategies, and mitigation of degradation. The NRC is reviewing new information submitted by GE on the safety significance of, and recommended inspections for, top guide and core plate ring cracking.

REFERENrfj 1601. Memorandum to C. Serpan from A. Chaffee, " Nuclear Reactor Regulation (NRR)

Input Into Research NUREG-0933 (WITS Item 9400213)," February 13, 1996.

1670. NRC Generic Letter 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors," July 25, 1994.

1671. Memorandum for A. Thadani from B. Sheron, " Staff Action Plan for the Resolution of Issues Associated With Boiling Water Reactor Internals Cracking," April 26, 1995.

O 06/30/96 A.D.001-2 NUREG-0933

(V RGA-002: REACTOR PRESSURE VESSEL FRACTURE TOUGHNESS DESCRIPTION As a result of the information provided by licensees in response to GL 92-01,2"'

Revision 1, issued in March 1992, the staff issued NUREG-15112'" and the Reactor i

Vessel Integrity Database (RVID). NUREG-1511 '" provides a summary of the 2

critical issues and regulatory requirements involved in RPV structural integrity and the status of each RPV with respect to the regulatory requirements. The RVID contains all the data that were submitted by licensees to demonstrate compliance with the regulatory requirements. Since licensees provide data during the life of their plants to demonstrate compliance with regulatory requirements, NUREG-15112'" and the RVID will require periodic upgrading.

In April 1995, the staff completed its evaluation of the Palisades pl ant compliance with the PTS Rule (10 CFR 50.61) and concluded that the Palisades RPV could be operated in compliance with the requirements of the PTS Rule through the plant's 14th refueling outage scheduled for late 1999. To extend the life of the Palisades RPV beyond 1999, the licensee has begun to plan for annealing of the RPV. The Palisades reactor vessel will be the first commercial nuclear vessel annealed in the U.S. to improve its fracture toughness. The staff will review the licensee's annealing plan prior to its implementation scheduled for the 1998 refueling outage. Prior to this anneal, the industry will perform demonstration (mv)anneals at the Marble Hill and Midland-2 sites. This activity was identified in an NRR memorandum '" to RES in February 1996.

2 10 CFR 50, Appendix G, and 10 CFR 50.61 establish requirements to prevent fracture of the RPV and require licensees to project the amount of embrittlement of RPV materials. As a result of the review of responses to GL 92-01,'"' the review of the Palisades PTS, and inspections conducted at CE offices in Windsor, Connecticut, several concerns related to RPV evaluations have been identified and are summarized as follows:

(1) It appears that licensees may not have been aware of, or considered, all relevant information and data in previous assessments of their RPVs; (2) The variability in copper and nickel chemical composition may be independent of weld heat number and is greater than previously recognized by the staff.

Based on the above findings, the staff conclu'ded that the most effective way to resolve the concern was through a supplement to GL 92-01'"' requiring licensees to collect all data relevant to their RPVs and, if there are data that had not been previously considered, to perform a reassessment of their RPVs. As a result of the data supplied in response to GL 92-01 "' and the Palisades PTS review, NRR requested RES to evaluate whether changes to the PTS Rule or RG 1.99"' are necessary.

Specific actions included in the generic action plan"" are: (1) issue Supplement 1 to GL 92-01'"*, (2) coordination with RES on RPV integrity issues; (3) hold an 06/30/96 A.D.002-1 NUREG-0933

NRC/ Industry workshop"on RPV issues; (4) review first and second round of responses to GL 92-01, '* Supplement 1; (5) issue Supplements 1 and 2 to NUREG-1511""; (6) issue Revisions 1 and 2 of the RVID; (7) observe industry annealing demonstrations; (8) review and evaluate the Palisades annealing plan; and (9) review the Palisades anneal.

1 CONCLUSION The staff's assessment"" of the impact of increased variability in chemistry on the RT,n value of PWR vessels indicated that there is no immediate cause for concern and that there is adequate time to perform a more rigorous assessment of the issue. GL 92-01 "

was held. Requests"I"" Supplement I was issued and an NRC/ Industry workshop""

for research on RPV inte RVID was issued (NRC Administrative letter 95-03'gr,ity were made by NRR and the

).

REFERENCES 841.

Regulatory Guide 1.99,

" Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, July 1975, (Rev. 1) April 1977, (Rev. 2) May 1988.

1601. Memorandum to C. Serpan from A. Chaffee, " Nuclear Reactor Regulation (NRR)

Input Into Research NUREG-0933 (WITS Item 9400213)," February 13, 1996 1603. Memorandum to A. Thadani from J. Strosnider, " Plan for Addressing Generic Reactor Pressure Vessel Issues," August 9, 1995.

1604. NUREG-1511,

" Reactor Pressure Vessel Status Report,"

U.S.

Nuclear Regulatory Commission, December 1994.

1605. Memorandum to A. Thadant from J. Strosnider, " Assessment of Impact of Increased Variability in Chemistry of the.RT,n Value of PWR Reactor Vessels," May 5, 1995.

1672. NRC Letter to All Holders of Operating Licenses or Construction Permits for Nuclear Power Plants (Except Yankee Atomic Electric Company, Licensee for the Yankee Nuclear Power Station),

" Reactor Vessel Structural Integrity,10 CFR 50.54(f) (Generic Letter 92-01)," February 28, 1992, (Rev. 1) March 6, 1992, (Rev. 1, Supplement 1) May 19, 1995.

1673. Memorandum for E. Beckjord from W. Russell, "NRR User Need Request for Support of Resolving Problem of Stress Corrosion Cracking of Reactor Vessel Internal Components," December 2,1994.

1674. Memorandum for D. Morrison from W. Russell, " Request for Research on Reactor Pressure Vessel Integrity," August 11, 1995.

1675. Memorandum to L. Shao from H. Mayfield, " Summary, NRC/NEI Workshop on Nuclear RPV Integrity," September 6, 1995.

1676. NRC Administrative Letter 95-03, " Availability of Reactor Vessel Integrity Database," August 4, 1995.

06/30/96 A.D.002-2 NUREG-0933

l 1

Q RGA-003: DRY CASK STORAGE OF SPENT FUEL DESCRIPTION Since 1986, several U.S. nuclear power plant licensees have installed independent spent fuel storage installations (ISFSIs), i.e., licensee-owned dry cask storage facilities; other licensees are also planning such installations. In recent years, licensees have encountered a number of problems during the fabrication, installation, and licensing of some of these ISFSIs and there has been an inconsistent level of performance by involved licensees and cask fabricators with respect to the use of dry cask storage of spent reactor fuel. Because of the anticipated increased industry effort in this area, the staff needed to fully l

understand the problems that occurred and take appropriate measures to reduce such problems in the future. Therefore, NMSS and NRR reviewed the lessons learned by all Offices and the Regions from past experience with ISFSIs and developed a plan'"* to resolve major concerns and problems; this plan was to be pursued in accordance with an M0V'"' between NRR and NMSS. This activity was identified in an NRR memorandum'"' to RES in February 1996.

The concern addresses dry storage of fuel that is several years old. Technical concerns have been addressed on a site-specific basis for existing facilities.

The action plan will improve guidance, enhance communications with industry and the public, and aid future applicants.

O The Action Plan'"* was developed to identify and resolve major concerns and i

problems in the area of dry cask storage of spent reactor fuel in ISFSIs.

Specific concerns encompassed by the plan include heavy load control, procedures for cask loading and unloading, failed fuel storage, change processes, inspection activities, and communications (internal and external). Concerns have been divided into the following categories: near-term technical; long-term technical; communications; and process issues. Actions included in the plan are: (1) review each general issue and identify the specific problems to be addressed; (2) develop corrective actions for each problem; and (3) implement the corrective actions.

CONCLUSION The following action plan items have been completed: cask trunnions; hydrostatic testing; cask weeping; and 10 CFR 72 reporting requirements. The Regions, NMSS, and NRR hold regular interface calls to discuss dry cask issues, training has been given to many of the affected staff, and NRC has established open communications with the newly-formed Nuclear Energy Institute Dry Cask Storage Working Group.

REFERENCES 1601. Memorandum to C. Serpan from A. Chaffee, " Nuclear Reactor Regulation (NRR)

Input Into Research NUREG-0933 (WITS Item 9400213)," February 13, 1996.

1608. Memorandum to J. Taylor from C. Paperiello and W. Russell, " Dry Cask Storage Action Plan," July 28, 1995.

06/30/96 A.D.003-1 NUREG-0933

_ - _. - ~_. - - - -...- -... - - _. - -.

1609. Memorandum to J. Taylor from W. Russell and R. Bernero, " Realignment of Reactor Decommissioning Program," March 15, 1995.

f e

O O

06/30/96 A.D.003-2 NUREG-0933

1 l

' C i

RGA-004: THERM 0-LAG FIRE BARRIERS DESCRIPTION In June 1991, NRR established a special team to review the safety significance and generic applicability of technical issues regarding the use of Thermo-La fire barriers. In April 1992, the special review team issued its final report 'g 2

which identified concens about fire endurance, combustibility, and ampacity derating. Subsequently, NRR prepared an Action Plan'"' to address the issues associated with Thermo-Lag and the NRC fire protection program. This plan included evaluation and resolution of generic Thermo-Lag fire barrier concerns regarding toxicity, construction and installation, fire endurance, ampacity derating, combustibility, seismic capabilities, and uniformity of materials. The plan also called for the staff to evaluate the special review team findings and public concerns, coordinate with NEI and licensees, conduct fire endurance and ampacity derating tests, and assess the NRC reactor fire protection program. This activity was identified in an NRR memorandum'"2 to RES in February 1996.

In response to Bulletin 92-01 "* and its Supplement, licensees with Thermo-Lag fire barriers established NRC-approved measures, such as fire watches, to compensate for possible inoperable fire barriers. The combination of compensatory measures and the defense-in-depth fire protection features provide an adequate level of fire protection until licensees implement permanent corrective actions.

Specific actions in the Action Plan'"' include: (1) the resolution of generic concerns raised by the special review team; and (2) resolution of plant-specific issues that emerge from the generic concerns.

CONCLUSION In June 1994, the Commission approved a staff recommendation to resolve Thermo-Lag concerns by requiring compliance with existing NRC requirements and to permit plant-specific exemptions, where justified. As of December 1995, the staff had issued several generic communications regardin Thermo-Lag fire 95-27,2"g ins 91-47,"" 91-79, *" 92-55,'"' 92-82,2"* g barriers # includin 94-22,2"' 94-34,2'"

94-86,2" 95-32,2"' and 95-49.'"* Two major items of the action plan remain to be completed: (1) mechanical properties tests; and (2) plant-specific fire test curve feasibility study.

On 10/03/95, NEI submitted to the NRC the NUCON International, Inc. Report 06VA764/04,

" Pyrolysis Gas Chromatography Analysis and Energy Dispersive Spectroscopy of Thermo-Lag Fire Barrier Samples." In its letter to the NRC, NEI stated that, on the basis of the tests, all samples (169 from 18 utilities representing 25 nuclear power plants) contained the constituents essential to fire barrier performance and the composition of the samples was consistent. The staff performed chemical composition tests and analyses at NIST which confirmed the results of the NEI analyses.

O Concerns about the reliability of information and data supplied by TSI prompted the staff to reassess previous technical conclusions and determine the extent to j

which the NRC or the nuclear industry relied on information supplied by TSI to 06/30/96 A.D.004-1 NUREG-0933

reach these conclusions. The staff identified and categorized the issues and previous conclusions and, on the basis of the results of the chemical analysis performed by NIST and NEI, concluded that additional action was not needed to reassess the issues or verify the conclusions. The staff continues to work with NIST to evaluate the feasibility of developing fire curves for rating fire barriers on the basis of representative nuclear power plant fire hazards rather than the fire curves specified in existing fire test standards.

REFERENCES 1601. Memorandum to C. Serpan from A. Chaffee, " Nuclear Reactor Regulation (NRR)

Input Into Research NUREG-0933 (WITS Item 9400213)," February 13, 1996.

1639. NRC Information Notice No. 91-47, " Failure of Thermo-Lag Fire Barrier Material to Pass Fire Endurance Test," August 6, 1991.

1640. NRC Information Notice 91-79,

" Deficiencies in the Procedures for Installing Thermo-Lag Fire Barrier Materials,"

December 6,

1991, (Supplement 1) August 4, 1994.

1641. NRC Information Notice 92-55, " Current Fire Endurance Test Results for Thermo-Lag Fire Barrier Material," July 27, 1992.

1642. NRC Information Notice 92-82, "Results of Thermo-Lag 330-1 Combustibility Testing," December 15, 1992.

1643. NRC Information Notice 94-22, " Fire Endurance and Ampacity Derating Test Results for 3-Hour Fire-Rated Thermo-Lag 330-1 Fire Barriers," March 16, 1994.

1644. NRC Information Notice 94-34, "Thermo-Lag 330-660 Flexi-Blanket Ampacity Derating Concerns," May 13, 1994.

1645. NRC Information Notice 94-86, " Legal Actions Against Thermal Science, Inc., Manufacturer of Thermo-Lag," December 22, 1994.

1646. NRC Information Notice 95-27, "NRC Review of Nuclear Energy Institute,

'Thermo-Lag 330-1 Combustibility Evaluation Methodology Plant Screening Guide,'" May 31, 1995.

1647. NRC Information Notice 95-32, "Thermo-Lag 330-1 Fl ame Spread Test Results," August 10, 1995.

l 1648. NRC Information Notice 95-49, " Seismic Adequacy of Thermo-Lag Panels,"

I October 27, 1995.

j 4

1681. Memorandum for J. Taylor from T. Murley, " Planned Actions to Address the Issues from the Office of Inspector General's Report on the NRC Staff's l

Review and Acceptance of Thermo-Lag 330-1 Fire Barrier Material," August 21, 1992.

1682. NRC Bulletin No. 92-01, " Failure of Thermo-Lag 330 Fire Barrier System to Maintain Cabling in Wide Cable Trays and Small Conduits Free from Fire j

Damage," June 24, 1992.

06/30/96 A.D.004-2 NUREG-0933

V RGA-004: THERM 0-LAG FIRE BARRIERS DESCRIPTION In June 1991, NRR established a special team to review the safety significance and generic applicability of technical issues regarding the use of Thermo-Lag fire barriers. In April 1992, the special review team issued its final report *'

which identified concerns about fire endurance, combustibility, and ampacity l

derating. Subsequently, NRR prepared an Action Plan "2 to address the issues 2

associated with Thermo-Lag and the NRC fire protection program. This plan included evaluation and resolution of generic Thermo-Lag fire barrier concerns regarding toxicity, construction and installation, fire endurance, ampacity derating, combustibility, seismic capabilities, and uniformity of materials. The plan also called for the staff to evaluate the special review team findings and public concerns, coordinate with NEI and licensees, conduct fire endurance and ampacity derating tests, and assess the NRC reactor fire protection program. This activity was identified in an NRR memorandum "2 to RES in February 1996.

2 In response to Bulletin 92-01 "' and its Supplement, licensees with Thermo-Lag 2

fire barriers established NRC-approved measures, such as fire watches, to compensate for possible inoperable fire barriers. The combination of compensatory measures and the defense-in-depth fire protection features provide an adequate level of fire protection until licensees implement permanent corrective actions.

Specific actions in the Action Plan *"2 include: (1) the resolution of generic i

concerns raised by the special review team; and (2) resolution of plant-specific issues that emerge from the generic concerns.

CONCLUSION In June 1994, the Commission approved a staff recommendation to resolve Thermo-Lag concerns by requiring compliance with existing NRC requirements and j

to permit plant-specific exemptions, where justified. As of December 1995, the staff had issued several generic communications regardin Thermo-Lag fire 95-27,2"g ins 91-47,"" 91-79, *" 92-55,2"2 92 - 8 2, '"' g94-barriers includin 94-86,*"#

95-32,2"' and 95-49. 2"' Two major items of the action plan remain to be completed: (1) mechanical properties tests; and (2) plant-specific fire test curve feasibility study.

On 10/03/95, NEI submitted to the NRC the NUCON International, Inc. Report 06VA764/04, " Pyrolysis Gas Chromatography Analysis and Energy Dispersive Spectroscopy of Thermo-Lag Fire Barrier Samples." In its letter to the NRC, NEI stated that, on the basis of the tests, all samples (169 from 18 utilitias representing 25 nuclear power plants) contained the constituents essential to fire barrier performance and the composition of the samples was consistent. The i

staff performed chemical composition tests and analyses at NIST which confirmed the results of the NEI analyses.

l A Concerns about the reliability of information and data supplied by TSI prompted the staff to reassess previous technical conclusions and determine the extent to which the NRC or the nuclear industry relied on information supplied by TSI to 06/30/96 A.D.004-1 NUREG-0933 l

l

reach these conclusions. The staff identified and categorized the issues and previous conclusions and, on the basis of the results of the chemical analysis performed by NIST and NEI, concluded that additional action was not needed to reassess the issues or verify the conclusions. The staff continues to work with NIST to evaluate the feasibility of developing fire curves for rating fire barriers on the basis of representative nuclear power plant fire hazards rather than the fire curves specified in existing fire test standards.

REFERENCES 1601. Memorandum to C. Serpan from A. Chaffee, " Nuclear Reactor Regulation (NRR)

Input Into Research NUREG-0933 (WITS Item 9400213)," February 13, 1996.

1639. NRC Information Notice No. 91-47, " Failure of Thermo-Lag Fire Barrier Material to Pass Fire Endurance Test," August 6, 1991.

1640. NRC Information Notice 91-79,

" Deficiencies in the Procedures for Installing Thermo-Lag Fire Barrier Materi al s, "

December 6,

1991, (Supplement 1) August 4, 1994.

1641. NRC Information Notice 92-55, " Current Fire Endurance Test Results for Thermo-Lag Fire Barrier Material," July 27, 1992.

1642. NRC Information Notice 92-82, "Results of Thermo-Lag 330-1 Combustibility Testing," December 15, 1992.

1643. NRC Information Notice 94-22, " Fire Endurance and Ampacity Derating Test Results for 3-Hour Fire-Rated Thermo-Lag 330-1 Fire Barriers," March 16, 1994.

1644. NRC Information Notice 94-34, "Thermo-Lag 330-660 Flexi-Blanket Ampacity Derating Concerns," May 13, 1994.

1645. NRC mformation Notice 94-86, " Legal Actions Against Thermal Science, Inc., Manufacturer of Thermo-Lag," December 22, 1994.

1646. NRC Information Notice 95-27, "NRC Review of Nuclear Energy Institute,

'Thermo-Lag 330-1 Combustibility Evaluation Methodology Plant Screening Guide,'" May 31, 1995.

1647. NRC Information Notice 95-32, "Thermo-Lag 330-1 Flame Spread Test Results," August 10, 1995.

1648. NRC Information Notice 95-49, " Seismic Adequacy of Thermo-Lag Panels,"

October 27, 1995.

1681. Memorandum for J. Taylor from T. Murley, " Planned Actions to Address the Issues from the Office of Inspector General's Report on the NRC Staff's Review and Acceptance of Thermo-Lag 330-1 Fire Barrier Material," August 21, 1992.

1682. NRC Bulletin No. 92-01, " Failure of Thermo-Lag 330 Fire Barrler System to Maintain Cabling in Wide Cable Trays and Small Conduits Free from Fire Damage," June 24, 1992.

06/30/96 A.D.004-2 NUREG-0933

I t

i i

1683. Memorandum for W. Russell from T. Murley, " Final Report - Special Review I

)

Team for the Review of Thermo-Lag Fire Barrier Performance," April 21, 1992.

1 l

l l

4 1

h l

i 06/30/96 A.D.004-3 NUREG-0933

RGA-005: RCS DRAINDOWN DESCRIPTION

{

On 09/17/94, Wolf Creek experienced a loss of RCS inventory while transitioning to a refueling shutdown. The event occurred when operators cycled a valve in the Train A side of the RHR system cross-connect line following maintenance on the valve, while at the same time establishing a flow path from the RHR system, Train B, to the refueling water storage tank for reborating train B. The fanure of the reactor operating staff to adequately control two incompatible activities resulted in transferring 9,200 gallons of hot RCS water to the RWST in 66 seconds. A study of the event was documented in AE0D/S95-01.2'" This activity was identified in an NRR memorandum" to RES in February 1996.

The Wolf Creek event represents a LOCA with the potential to consequentially fail all the ECCS pumps and bypass the containment. Another important feature of this event is the short time available for corrective action. Based upon calculations by the licensee and the staff, it is estimated that if the draindown had not been isolated within 3 to 5 minutes, NPSH would have been lost for all ECCS pumps, and core uncovery would have followed in about 25 to 30 minutes. This event represented a vulnerability in PWRs that was not previously recognized.

2 Plan "' was developed to collect and evaluate information from O

An Action licensecs regarding plant-specific system configurations and vulnerabilities to draindown events. Specific actions included in the plan were staff issuance of:

(1) an IN to alert licensees to the Wolf Creek event; and (2) a GL requesting all PWR licensees to provide information on draindown vulnerabilities and the measures that have been implemented to diminish the probability of a draindown.

CONCLUSION The staff performed an evaluation of the probability of event initiation and of the conditional core damage probability. The resultant low value of the core damage probability along with licensee awareness of the failure scenario made the risk from the Wolf Creek event small and IN 95-03 "' was issued.

2 REFERENCES 1601. Memorandum to C. Serpan from A. Chaffee, " Nuclear Reactor Regulation (NRR)

Input Into Research NUREG-0933 (WITS Item 9400213)," February 13, 1996.

1649. NRC Inforrration Notice 95-03, " Loss of Reactor Coolant Inventory and Potential Loss of Emergency Mitigation Functions While in a Shutdown Condition," January 18, 1995, (Supplement 1) March 25, 1996.

1677. AE0D/S95-01, " Reactor Coolant System Blowdown at Wolf Creek on September 17, 1994," Office for Analysis and Evaluation of Operational Data, U.S.

Nuclear Regulatory Commission, March 1995.

1678. Memorandum to A. Thadani from R. Jones, " Proposed Action Plan for-the

' Wolf Creek Draindown Event,'" November 20, 1995.

06/30/96 A.D.005-1 NUREG-0933

RGA-006: SRP REVISION QESCRIPTION The SRP Update and Develcpment Program (SRP-UDF) was established"' in 1991 to update the SRP" for use in reviewing future reactor design applications. The revised SRP" incorporats:: changes in the regulation of the nuclear power industr 161,y that have occurred since the last SRP revision in 1981. In SECY the staff discussed, in part, the revision effort for the SRP and committed to produce supplements to the 19 ell SRP in parallel with the corduct of future reactor design reviews.

The SRP Revision Action Plan '*' deals with the development of draft revisions for 2

all sections, except Chapter 7, and the development of new SRP sections to cover review areas that are supported by established staff positions or are fully addressed in the evolutionary reactor design reviews. The draft revisions will incorptrate recommended changes identified in the review of generic regulatory documents and NRR staff SERs for evolutionary LWR designs. Specific tasks included in the Action Plan are:

(1) Identify established staff positions and new regulatory requirements from a review of generic regulatory documents issued since the last SRP revision and from a review of NRR staff safety evaluation reports for evolutionary LWR designs; (2) Prepare a side-by-side comparison of the SRP-cited version of codes and standards vs. the current version of the standard; (3) Prepare draft revisions of the current SRP sections to incorporate the changes recommended; (4) frepare new draft SRP sections that are supported by established staff positions or are fully addressed in the evolutionary design reviews; (5) Automate the SRP to make future res !sions and accessibility easier to accomplish; (6) Maintain the progran, data base to reflect new staff positions and requirements.

This activity was identified in an NRR memorandum"' to RES in February 1996.

[QELUSION NRR has established the SRP-UDP to update the SRP for use in the review of future reactor applications to reflect existing agency requirements and guidance and to add new review criteria to accommodate future designs. An automated version of the current SRP has been developed and is operational.

06/30/96 A.D.006-1 NUREG-0933

REFERENCES 11.

NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," U.S. Nuclear Regulatory Commission, (1st Edition) November 1975, (2nd Edition) March 1980, (3rd Edition) July

1981, 1601. Memorandum to C. Serpan from A. Chaffee, " Nuclear Reactor Regulation (NRR)

Input Into Research NUREG-0933 (WITS Item 9400213)," February 13, 1996.

1606. SECY-91-161, " Schedules for the Advanced Reactor Reviews and Regulatory Guidance Revisions," May 31, 1991.

1607. Memorandum for W.

Russell from F.

Gillespie, " Action Pl an for the Development of Draft SRP Revisions in the SRP-UDP," May 17, 1994.

1679. Memorandum for the Chairman from J. Taylor, " Commercial Contract for Technical Assistance to Support the Standard Review Plan Update and Development Program," November 18, 1991.

1680. Memorandum for J. Taylor from I. Selin, " Commercial Contract for Technical Assistance to Support the Standard Review Plan Update and Development Program," December 13, 1991.

O O

06/30/96 A.D.006-2 NUREG-0933

m RGA-007: PRA IMPLEMENTATION PLAN The NRC has been making use of PRA technology to varying degrees in its regulatory activities since the issuance of WASH-1400.2' Prior to 1991, this had been an ad hoc application, depending on the availability of expertise in various technical groups. Since 1991, there have been a number of high level studies within NRC that focused on the status of PRA use and its role in the regulatory process. Collectively, the findings and recommendations from these studies support the view that there is a need for increased emphasis on PRA technology applications.

For the full value of the NRC'S investment in risk assessment methodology to be achieved, it is important that consistent high-level agency guidance be provided on the appropriate use of PRA. To this end, in November 1993, the Office Directors of NRR, AE0D, NMSS, and RES proposed to take the initiative in providing guidance on coordination and expectations for PRA efforts.

Specifically, they proposed to develop an integrated plan for the staff's risk assessment and risk management practices.2*" In August 1994, the staff submitted SECY-94-219'"* for the Commission's information. On 03/30/95, the staff submitted SECY-95-079'"' and briefed the Commission on the subject on 04/05/95. On 05/18/95, the staff forwarded SECY-95-126'"' to the Commission and a final PRA Policy Statement was published in the Federal Register on 08/16/95.

O An Action Plan'" was developed to describe the process for the staff to use PRA methods and technology in the NRC's effort toward a risk-informed regulatory approach; this plan encompasses methods development, pilot applications, and staff training. The plan will be used to ensure timely and integrated agency-wide effort that is consistent with the PRA Policy Statement and is meant to improve the regulatory process by developing state-of-the-art PRA tools that will expand the use of PRA technologies in making regulatory decisions; the plan is not intended to correct safety problems at licensed facilities.

The PRA Implementation Plan includes activities for NRR, RES, AE00, and NMSS staff to increase the use of PRA methods in all regulatory matters. NRR focuses on the PRA applications in reactor regulations, the development of standard I

review plans, the pilot programs to use PRA technology in specific regulatory initiatives, events assessment, and working with regions on risk-informed inspections. RES focuses on the IPE/IPEEE reviews, PRA method and quality, and the development of PRA regulatory guides for the industry. AE00 focuses on risk-informed trends and patterns analysis, reliability data for PRA applications, and staff training. NMSS focuses on using PRA in high and low level waste issues. The detailed actions are described in the PRA Implementation Plan.

On 11/17/95, a memorandum was forwarded to senior NRR management providing I

additional guidance on implementing the Commission's PRA Policy Statement and i

managing tasks contained in the PRA Implementation Plan. As a result of this j

memorandum, several additional action plans were expected to be developed for i

individual line items in the PRA Implementation Plan. In addition, more detailed O-information concerning PRA Implementation Plan activities will be collected so

)

that more accurate and timely status of all activities can be maintained in the ongoing PRA Implementation Plan. On 11/27/95, the staff issued SECY-95-280" to i

06/30/96 A.D.007-1 NUREG-0933 2

provide a general structure to ensure consistent and appropriate application of PRA methods and to outline a process for developing guidance and standards. On 11/30/95, Chairman Jackson issued a memorandum requesting the staff to develop action plans and timetables to provide better focus and accelerate NRC's risk-informed re ulatory effort.

memorandumg' to Chairman Jackson onIn response to this request, the EDO forwarded a 01/03/96 which described the staff action plan for utilizing PRA in reactor-related activities, including the PRA pilot programs and the accelerated milestones for the development of regulatory guidance documents.

RGA-007.1.2(D): GRADED OVALITY ASSURANCE DESCRIPTION This task called for the preparation of staff evaluation guidance and regulatory guidance for industry implementation for the grading of QA practices commensurate with the safety significance of the plant equipment. The development of this guidance will be based on staff reviews of regulatory requirements, proposed changes to existing practices, and assessment of the actual programs developed by the three volunteer utilities implementing graded QA programs.

10 CFR 50 Appendices A and B require QA programs that are commensurate (or consistent) with the importance to safety of the functions to be performed.

However, the QA implementation practices that have evolved have often not been graded. In the development of implementation guidance for the maintenance rule, a methodology to determine the risk significance of plant equipment was proposed by the industry (NUMARC 93-01). During a public meeting on 12/16/93 the staff suggested that the industry could build on the experienca gained from the maintenance rule to develop implementation methodologies for graded QA. The staff had numerous interactions with NEI during 1994 as the graded QA concepts were discussed and the initial industry guidelines were developed and commented on.

In early 1995, the licensees of Grand Gulf, South Texas, and Palo Verde volunteered to work with the staff. The staff has reviewed the licensee developmental graded QA efforts. This activity was identified in an NRR memorandum **

• to RES in February 1996.

The goal of the Action Plan'" is to utilize the lessons learned from the 3 volunteer licensees to modify staff-developed draft guidance to formulate regulatory guidance on acceptable methods for implementing graded QA. The staff will develop a regulatory guide, a revision to Chapter 17 of the SRP, and a reactive inspection procedure (IP) for graded QA. An inter-office team will be established to prepare the regulatory guidance documents and test their implementation during the evaluation of volunteer plant activities.

Existing regulations provide the necessary flexibility for the development and implementation of graded QA programs. The staff will issue a NUREG report regarding the lessons learned from the volunteer plant implementations.

Additional regulatory guidance will be issued to either disseminate staff guidance or endorse an industry approach. Planned guidance for the staff will involve an evaluation guide for application to the volunteer plants, the lessons learned report, training sessions and public workshops, SRP revisions, and inspection guidance in the form of a reactive IP. The staff is evaluating the 06/30/96 A.D.007-2 NUREG-0933

O appropriate mechanism for inspections of the risk significance determination b

aspects of graded QA programs.

The safety benefits to be gained from a graded QA program could be significant since both NRC reviews and inspections and the industry's quality control resources would be focused on the more safety significant plant equipment and activities. Secondarily, cost savings to the industry could be realized by avoiding the dilution of resources expended on less safety significant issues.

CONCLUSION A draft evaluation guide for NRC use has been prepared for application to the volunteer piants implementing graded QA programs. The staff will utilize the guide for the review of the volunteer plant graded QA programs. The guide and the staff's proposed interaction framework were expected to be transmitted to the three volunteer licensees for comments.

REFERENCES 1601. Memorandum to C. Serpan from A. Chaffee, " Nuclear Reactor Regulation (NRR)

Input Into Research NUREG-0933 (WITS Item 9400213)," February 13, 1996.

1630. SECY-94-219, " Proposed Agency-Wide Implementation Plan for Probabilistic Risk Assessment (PRA)," August 19, 1994.

1631. SECY-95-079, " Status Update of the Agency-Wide Implementation Plan for

(

Probabilistic Risk Assessment," March 30, 1995.

1632. SECY-95-126, " Final Policy Statement on the Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities," May 18, 1995.

1633. SECY-95-280, " Framework for Applying Probabilistic Risk Analysis in Reactor Regulation," November 27, 1995.

1634. Memorandum to Chairman Jackson from J. Taylor, " Improvements Associated with Managing the Utilization of Probabilistic Risk Assessment (PRA) and Digital Instrumentation and Control Technology," January 3,1996.

1635. Memorandum for J. Taylor from T. Murley, et al., " Agency Directions for Current and Future Uses of Probabilistic Risk Assessment (PRA)," November 2, 1993.

J o

06/30/96 A.D.007-3 NUREG-0933

a a_

2.

..,-s%._.m.wmm.s a

..m..asemae.

a.m.,,

e,aaha4-um A4n-,.u.a.A4

.#.w a-+-

.4 e4.EAa-e.

a

-44A 4 424me.mm 4 A._ A & & M 4_A_4-aa4

.46a-..w_A.4_,J-4-4 au-A.

6 d

l l

l

)

I O

I I

9

i I

1 4

l l

APPENDIX E GENERIC COMMUNICATION AND COMPLIANCE ACTIVITIES z

l i

This appendix documents those generic communication and compliance activities (GCCA) completed by NRR that did not meet the criteria for designation as generic issues (GI), but were important enough to require the issuance of Information i

Notices (IN) and/or Generic Letters (GL) to licensees. The plan for documenting closed GCCAs was delineated in SECY-96-107.'"

}

}

l 1

e d

I I

06/30/96 A.E-1 NUREG-0933

GCCA-0001: ASSESSMENT OF CONDITION OF SAFETY-RELATED STRUCTURES AND CIVIL ENGINEERING FEATURES TAC No.: M87093

Contact:

R.A. Benedict Descriotion: Nuclear power plant structures are designed to withstand low-probability natural phenomena and reactor accident loadings and are constructed utilizing stringent quality control requirements. They are robust and have not i

been subjected to the low-probability challenges for which they were designed except on two occasions: the accident at TMI-2 and the fierce wind luadings imposed by Hurricane Andrew on the structures of Turkey Point 3 & 4. Structures subjected to the loadings from these events withstood the loads without appreciable damage. However, information on the failures of non-nuclear structures, such as highway bridge-decks and parking garages, indicates that the age-related degradation of well-designed and properly constructed concrete structures could weaken them sufficiently to cause them to fail without being subjected to abnormal loadings. Several incidents of age-related degradation of nuclear structures have been reported.

The objectives of this study were to: (1) review the known information on the degradation of structures and assess their conditions with respect to their safety functions; (2) make observations as to whether these safety functions are maintained for the life of the plant; and (3) provide information that could be useful for the improved design and construction of structures of future reactors.

Oriainatina Document.:

NUREG/CR-4652,

" Concrete Component Aging and Its Significance Relative to Life Extension of Nuclear Power Plants," September 1986.

Reaulatory Assessment: Six vintage plants were visited to collect information for assessing the existing condition and past performance of structures and civil engineering features such as Seismic Category I buildings, tanks, cable trays, conduit and equipment supports, underground structures, water intake structures and anchorages, and fuel racks. Structural components included reinforced concrete, structural steel, and masonry walls.

The review concluded that most of the civil / structural plant features have performed very well.

Some structures / components thowed signs of aging degradation. The 11 degradation categories were listed and rated for each plant.

The ratings were not judgments about the current overall safety of these specific plants.

For license renewal applications, the ratings provide guidance for identifying the types of degradation that may require detailed review during the license renewal process and note the desirability of reguli inspections and maintenance of particular structures and equipment. For future plants, the ratings provide guidance for identifying the types of potential degradation that need to be addressed during the licensing process. The report presents the results of the staff's study of the six plants; it does not indicate a need for immediate action with respect to operating plant safety.

Resolution: Issuance of NUREG-1522, " Assessment of Inservice Conditions of Safety-Related Nuclear Plant Structures."

Completion Date: 08/28/95 06/30/96 A.E-2 NUREG-0933

l t

GCCA-0002: ENVIRONMENTAL LICENSING AND REGULATORY CONCENTRATIONS IN BUILDING

\\

WAKES TAC No.: M87875

Contact:

C.V. Hodge Descriotion: GDC 19 of 10 CFR 50, Appendix A, sets forth the requirements for control rooms at nuclear power plants. This criterion states that "[a]dequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures is excess of 5 rem." Computational means for assessment of licensee compliance with this requirement are lacking.

Oriainatina Document: GI 83, " Control Room Habitability."

Reaulatory Assessment: Beginning in the mid-1980s, RES sponsored development of i

a computer code on diffusion in the vicinity of buildings. PNL completed development of a computer code ARCON95, described in NUREG/CR-6331, " Atmospheric Relative Concentrations in Building Wakes," May 1995. This code uses hourly j

meteorological ds. and recently developed methods for estimating dispersion in the vicinity of Lildings to calculate relative concentrations at control room l

air intakes that would be exceeded no more than 5% of the time. These l

concentrations are calculated for averaging periods ranging from one hour to 30 l

days in duration. Relative concentrations calculated by ARCON95 are significantly i

lower than concentrations calculated using the currently accepted procedure when winds are less than two meters /second. For higher wind speeds, ARCON95 calculates about the same concentrations as the current procedure.

Resolution: EMCB is tracking RES activity on GI-83 on control room habitability l

but has no direct activities associated with the GI; therefore, the TAC No. was closed.

Completion Date: 07/21/95 GCCA-0003: RRG. 50.54(P) GUIDA!{GI l

Contact:

J.W. Shapaker Descriotion: On 01/04/93, the E00 established a Regulatory Review Group (RRG) to conduct a review of power reactor regulations and related processes, programs, and practices. One resulting RRG recommendation was to change the current l

practice that enables licensees to make changes to their security plans without prior NRC approval, i.e., using the provisions of 10 CFR 50.54(p).

Oriainatina Docum_ent: RRG Final Report, Volume 2 - Regulations, Section 2.3, Position Paper 2.3.18, " Security," dated August 1993.

Reaulatory Assessment: The plan developed by the staff for implementing the RRG recommendation was not to change the regulations, but +o clarify the process by providing revised screening criteria that would en:

e consistency of security plan changes without prior NRC approval.

Use of the revised screening criteria would allow licensees to reduce certain l

commitments that have exceeded regulatory requirements or published guidance if 06/30/96 A.E-3 NUREG-0933

___y

_-r y.-.

the overall effectiveness of the plan is not reduced. Each issue is reviewed against the overall assurance levels contained in the plan and not against specific individual changes. Latitude has always existed in that improvements in one area of the program may offset reductions in other areas. Overall assurance levels of the plans must be maintained and this clarification is not intended to reduce plan commitments to levels less than the overall high-assurance objectives stated in 10 CFR 73.55(a).

NRC has expected that licensees would judiciously make the proper determination regarding 10 CFR 50.54(p) changes and implement those changes as permitted by the regulations. This position was the original intent of the Commission and remains so today. The NRC believes that, with the use of the revised screening criteria and expertise of the licensee staff, licensees should implement changes made pursuant to 10 CFR 50.54(p) without prior NRC approval.

~

Resolution: Issuance of GL 95-08, "10 CFR 50.54(p) Process for Changes to Security Plans Without Prior NRC Approval," dated 10/31/95.

Comoletion Date: 10/31/95 GCCA-0004: RELOCATION 0F SELECTED TS REQUIREMENTS RELATED TO INSTRUMENTATION (GL)

Contact:

J.W. Shapaker Descriotion: Licensaes that have not converted, or are not in the process of converting, to the improved STS may request a license amendment to relocate selected instrumentation requirements from their TS.

This line-item TS improvement was developed in response to TS amendments proposed by licensees and NRC TS improvement initiatives.

Oriainatina Documents: NRC Region III Morning Report 3-95-0019, dated 02/22/95; 10 CFR 50.72, Event Notification 28926, dated 06/11/95.

Reaulatory Assessment: Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TS as part of the license.

In 10 CFR 50.36, the Commission establishea the regulatory requirements related to the content of TS; however, the regulation does not specify the particular requirements to be included in TS. The NRC developed criteria, as described in the " Final Policy Statement" (Federal Register Notice 58 FR 39132), to determine which of the design conditions and associated surveillances should be located in the TS as limi'ing conditions for operation. Four criteria were subsequently t

incorporated into the regulations by an amendment to 10 CFR 50.36 (Federal Register Notice 60 FR 36953):

(1) installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the RCPB; (2) a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; 06/30/96 A.E-4 NUREG-093'

O(

(3) a structure, system, or component that is part of the primary success path

")

and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (4) a structure,

system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

Implementation of these criteria may cause some requirements to be moved out of existing TS to documents and programs controlled by licensees. In this regard, GL 95-10 addresses the relocation of selected TS requirements related to instrumentation as a result of applying the 10 CFR 50.36 criteria; explicit guidance is given to licensees for submitting a proposed license amendment to accomplish this.

Resolution:

GL 95-10,

" Relocation of Selected Technical Specifications Requirements Related to Instrumentation," dated 12/15/95.

Completion Date: 12/17/95 GfCA-0005: BWR - SCRAM SOLEN 0ID PILOT VALVE PROBLEMS TAC No.: M90285

Contact:

D.L. Skeen A

I 1

Descriotion: Slow scram times were noted at some BWR plants as a result of the V

fluoroelastomer material used in ASCO HV-176-816 (T-ASCO) scram solenoid pilot valves (SSPVs). An apparent change in the material supplied to ASCO by a sub-supplier caused the material to soften and allowed the SSPV to stick immediately after being deenergized. The sticking SSPV caused a small delay in the scram time of the affected control rod.

Oriainatina Document: Grand Gulf Event Notification 26996, dated 03/27/94.

Reaulatory Assessment: A relatively few number of plants were affected. All four of the BWR-6 plants, one BWR-5 plant and 2 BWR-4 plants use the T-ASCO valves in their scram systems. However, only two of these plants have reported slow scram times that were attributed to the fluoroelastomer. Discussions with GF indicated that only the BWR-6 plants are more likely to be affected because of the faster scram requirements for that vintage plant. The delays seen were in the 10 to 400 millisecond range. Even though only two plants reported slow times, the other plants were made aware via the nuclear network entries made by those two plants and GE SIL No. 591, " Delayed SCRAM Solenoid Pilot Valve Operation," issued on 05/12/95.

Resolution: Based on the small number of potentially affected plants, the improved specifications for the elastomers developed by GE and ASCO as of April 1995, and the fact that GE identified the potential problem to licensees in a SIL, the Events Assessment Panel canceled the development of an information notice on C,7/18/95.

Completion Date: 07/18/95 06/30/96 A.E-5 NUREG-0933

GCCA-0006: SHIFT STAFFING ISSUE FOLLOWUP TAC No.: M91163

Contact:

N.K. Hunemuller

==

Description:==

On 11/26/91, the NRC issued IN 91-77, " Shift Staffing at Nuclear Power Plants," to alert licensees to the problems that could result from inadequate controls to ensure that shift staffing is sufficient to accomplish all functions required by an event. However, after IN 91-77 was issued, event follow-up inspections indicated that problems involving shift staffing and task allocation continued to occur. As a result, the NRC continued with further research in this area. This research included an RES project to address the adequacy of minimum shift staffing levels through a shift staffing study encompassing all licensee staff initially needed during an event.

Oriainatina Document: NRR Task Action Plan, " Nuclear Power Plant Shift Staffing,"

dated 04/13/95.

Reaulatory Assessme_01: The licensees surveyed generally staffed to levels greater than those required by either the regulations or their plant-specific TS for both licensed and non-licensed personnel. Nevertheless, the results of the research project provide several insights into areas which could impact the ability to accomplish safety functions following an event.

Resolution: IN 95-48, "Results of Shift Staffing Study," dated 10/10/95, was issued to inform licensees of the results of the NRC's study conducted as part of the RES project to address the adequacy of minimum shift staffing levels at nuclear power plants.

Comoletion Date: 10/10/95 GCCA-0007: LESSONS LEARNED FROM OPERATIONAL SAFEGUARDS RESPONSE EVALUATIONS TAC No.: M91231

Contact:

E.J. Benner

==

Description:==

At 6:53 a.m. on 02/07/93, an intruder drove into the TMI site entrance, continued past the guard house, and crashed through Gate I of the j

protected area. The vehicle proceeded to crash through a turbine building roll-up i

door and came to a st'op 63 feet inside the turbine building, enveloped by the roll-up door. Control room personnel responded by implementing emergency response procedures, including locking control room fire doors, and classifying the event as a Site Area Emergency (SAE) at 7:05 a.m. Security staff responded by posting security personnel to intervene at pre-designated vital areas, confirming vital area integrity and, with the aid of offsite responders, conducting an assessment of and search for the intruder. TMI security personnel found and apprehended the unarmed intruder at 10:57 a.m. The intruder was located at the bottom of the turbine building in a small space under condenser piping and of fered no resistance. The intruder was questioned onsite by Pennsylvania State Police then escorted offsite in custody. The U.S. Army explosives ordinance disposal unit completed a detailed search confirming that no explosives were present. Upon visually inspecting plant equipment, verifying plant parameters, and confirming that safety systems were available, the licensee terminated the SAE at 4:25 p.m.

Oriainatino Document: Event Notification 25035, dated 02/07/93.

06/30/96 A.E-6 NUREG-0933

O Reaulatory Assessment: Due to the vehicle intrusion event at TMI, the EDO assigned NRR, NMSS, AE00, RES, and Region I responsibility for taking generic actions resulting from the investigation of the unauthorized forced entry of the protected area at TMI on 02/07/93. An action plan was developed under TAC No.

M40031 to provide generic guidance on unauthorized forced entry.

The action plan determined that lessons learned from the Operational Safeguards Response Evaluations should be issued to licensees in the form of an IN; TAC No.

M91231 was for development of the IN. Because of the sensitive nature of the information, it was subsequently determined that an IN was an inappropriate method of informing licensees. Instead, the decision was made to send individual letters to all plant security managers with the safeguards-classified information attached.

Resolution: Issuance of individual letters to all plant security managers with safeguards-classified information attached. (See Accession No. 9509260227 for a sample letter.)

Comoletion Date: 12/12/95 GCCA-0008: AIR ENTRAINMENT IN TERRY TURBINE LUBRICATING CONTROL OIL SYSTEM TAC No.: M91307

Contact:

D.L. Skeen

==

Description:==

Entrainment of air into the oil system that serves the dual purpose O

of lubrication of the turbine bearings and hydraulic speed control was identified Vj as a problem for AFW systems in PWRs and RCIC systems in BWRs. The HPCI systems in BWRS were not susceptible to this problem because they are equipped with separate oil pumps that preclude the air entrainment. The air entrainment presents a potential problem of overheating of a turbine bearing or turbine speed fluctuations.

Oriainatina Document: Pilgrim Event Notification 27621, dated 08/03/94.

Reaulatory Assessment: Air entrainment could present a potential problem to both AFW and RCIC systems. The vendor stated that a 4-hour test run at the factory and initial start-up testing at the plant usually identify the problem and a larger drain line and/or a vent at the bearing housing can be installed to prevent any further problems with air entrainment. Despite the vendor claims, reports from at least three plants experiencing air entrainment problems prompted the NRC to issue an IN.

Resolutiqn: IN 94-84, " Air Entrainment in Terry Turbine Lubricating 011 System,"

was issued on 12/02/94 to address the air entrainment issue, while the need for a GL was being considered. Subsequent eveals involving binding of the governor valve stem and blockage of the turbine drain pots prompted the NRC to consider Terry Turbine reliability as a whole and TAC No. M92636 was opened to track the long-term follow-up of Terry Turbine activities. Since the air entrainment concern was included in the long-term follow-up, TAC No. M91307 was closed based on a memorandum from G. Holohan to 3. Grimes dated 07/21/95.

(

Comoletion Date: 07/21/95 06/30/96 A.E-7 NUREG-0933

GCCA-0009: WRONG REPLACEMENT PARTS RELIEF VALVES AND REFUELING MAST TAC No.: M91399

Contact:

J.L. Birmingham Descriotion: This proposed IN discussed three instances in which the installation of incorrect replacement parts resulted (or may have resulted) in equipment damage or improper performance. The instances occurred at River Bend on 09/21/94, at Cooper on 08/30/94, and at ANO-2 in July 1990.

Oriainatina Document: Memorandum from W.P. Ang to A.E. Chaffee, signed 05/15/95, propos.ing that an IN be issued to address the installation of incorrect replacement parts.

Reaulatory Assessment: The probable adverse. consequences of using incorrect replacement parts in safety or non-safety-related equipment is a concern that is generally well understood within the nuclear industry. Although the concern for the use of incorrect parts in equipment is valid, the information in the proposed notice did not provide a new or different understanding of the issue. N 'oncern that licensees use correct parts during equipment repair or replac+., '.s best addressed by licensc.e audit programs and by NRC inspection of licensee programs for maintenance. Therefore, an IN on this issue is not necessary at this time.

Resolution: The need for the proposed IN was discussed with representatives of Region IV (the originating group) and agreement was reached that the proposed notice was not needed at this time. The proposed notice was cancelled 07/11/95 via an E-mail message from W.P. Ang, Region IV.

Comoletion Date: 07/20/95 GCCA-0010: SPENT FUEL POOL OVERFLOW INTO VENTILATION SYSTEM TAC No.: M91400

Contact:

N.K. Hunemuller

==

Description:==

A draft Ih from Region II discussed the results of follow-up reviews of an event at Brunswick-2 when water from the spent fuel pool overflowed into the ventilation system and drained onto several elevations of the reactor building floor.

Oriainatina Document: Memorandum from B. A.

Boger to B.K. Grimes, " Proposed Information Notice - Spent Fuel Pool Overflow," dated 11/21/94.

Reaulatory Assessment: This single plant-specific event did not raise to the level of concern requiring the issuance of an IN. NRR Task Action Plan, " Generic Spent Fuel Storage Pool," addresses spent fuel pool safety concerns.

H.esol ution: After management review, the IN was cancelled based on the 07/21/95 memorandum from G.M. Holahan to B.K. Grimes, " Review of Generic Cummunications and Compliance Activities." The proposed IN failed to meet the threshold for regulatory action based on its safety significance and the number of affected plants. However, spent fuel pool safety concerns are being addressed in the NRR Task Action Plan, "Ganeric Spent Fuel Storage Pool."

Comoletion Date: 07/21/95 06/30/96 A.E-8 NUREG-0933

GCCA-0011: IPEEE FOR SEVERE ACCIDENT VULNERABILITIES TAC No.: M91401

Contact:

J.W. Shapaker

==

Description:==

In 1991, GL 88-20, Supplement 4 was issued requesting all licensees to perform an IPEEE to fid plant-specific vulnerabilities to severe accidents i

caused by external events, including seismic events, and report the results to the NRC. A companion document, NUREG-1407, provided procedural and submittal guidance for the IPEEE. Review level earthquakes (RLEs) and the review scope were defined by the staff for all U.S. sites; plants in the central and eastern U.S.

were assigned to appropriate review categories (plant bins) primarily according to a comparison of available seismic hazard results.

The hazard results used in the binning process included those published in 1989 by the LLNL (NUREG/CR-5250) and the EPRI (NP-6395-D). NRC established the bins because of the larga inherent uncertainties in the probabilistic estimation of seismic hazard. Using this approach, the staff campared the relative seismic i

hazard of the 69 central and eastern U.S. plant sites, and assigned each plant i

to one of four bins for the seismic margins method (Reduced-Scope, 0.39 Focused-Scope, 0.3g Full-Scope, and 0.59 bin).

In 1994, based on a re-elicitation of LLNL ground-motion and seismicity experts, the staff published revised seismic hazard results in NUREG-1488. The new LLNL mean hazard estimates were lower than the 1989 LLNL results but higher than the EPRI estimates. In a letter from W. Rasin (NEI) to A. Thadani (NRC) on 04/05/94, O

NEI advocated that most focused-scope plants should instead perform reduced-scope studies as part of the seismic IPEEE, based on the revised hazard estimates. NEI also stated that each licensee is responsible for proposing the most cost-effective program to satisfy the seismic IPEEE request consistent with the level of seismic hazard at the specific site. As a result, seven licensees informed the NRC of their intent to revise their IPEEE commitments.

These developments prompted the NRC to systematically revisit the seismic IPEEE 1

program rather than deal with each licensee individually. In IN 94-32, the staff stated that it would review LLNL's revised seismic hazard estimates and determine l

the appropriateness of revising the seismic IPEEE scope.

NRC contracted with Ener9y Research, Inc. (ERI) to do a seismic revisit study to determine whether consideration of the new LLNL seismic hazard estimates would:

(1) significantly change the original binning results; and (2) warrant adjusting the seismic scope and guidelines of the seismic IPEEE review. The latter effort would also require the determination of how the scope should be modified and the justification of such modifications. ERI completed the study and submitted two reports (ERI/NRC 94-50?. and ERI/NRC 94-504). The staff subsequently held a public workshop to discuss these reports, present comments from a peer review group, determine issues to be addressed, and solicit public input for developing the staff position on the seismic scope modification.

Oriainatina Document: IN 94-32, " Revised Seismic Hazard Estimates," dated 04/29/94.

Reaulatory Assessment: The NRC staff evaluated the ERI reassessment reports, the g

peer review group comments, the NEI white paper and comments received at and after the workshop. The staff concluded that: (1) licensees may use the revised 06/30/96 A.E-9 NUREG-0933 3

LLNL seismic hazard estimates instead of the 1989 LLNL seismic hazard estimates in the seismic PRA; and (2) the scope of the seismic IPEEE may be modified for all focused-scope and full-scope plants by eliminating the need to calculate the capacity of certain generally rugged components, or certain site effects that would not be significant sources of contributors to seismic severe accident risk or would not result in cost-beneficial improvements. The justification for this reduction in the seismic review scope is that the perceived seismic hazard estimates and associtted risks have decreased. However, the examination process for the modified seismic IPEEE remains the same as the process described in Supplement 4 to GL 88-20 and NUREG-1407.

Resolution: Issuance of GL 88-20, Supplement 5, " Individual Plant Examination of External Events for Severe Accident Vulnerabilities," dated 09/08/95.

Completion Date: 09/08/95 GCCA-0012: SURRY VENTILATION FILTER ISSUE TAC No.: M91438

[ontact: R.A. Benedict

==

Description:==

Cleaning the secondary side of a steam generator was performed at Surry Unit 2 using chemicals which, when vented through the charcoal adsorbers of the ventilation system, reduced significantly the iodine-reraval efficiency of the charcoal. The licensee was not aware that some of the chemicals used could degrade the charcoal, Oriainatina Document: Oral report from G. Hubbard, followed by NRC Inspection Report 50-280/94-21.

Reaulatory Assessment: Exposure of charcoal to chemical compounds can result in a degradation of the charcoal. In the event of a nuclear accident, the degraded charcoal in the ESF filtration systems may perform at an efficiency significantly less than that assumed in the DBA dose assessments in the staff's safety evaluation. Chemical cleaning is performed on a frequent basis at most PWRs, either every or every other refueling outage. The IN is needed to inform the industry of the potential problem associated with chemical cleaning of components.

Resolution: Issuance of IN 95-41, " Degradation of Ventilation System Charcoal Resulting from Chemical Cleaning of Steam Generators," dated 09/22/95.

Comoletion Date: 09/30/95 GCCA-0013: COMMON MODE FAILURE OF COPES VOLCAN PORVs TAC No.: M91446

Contact:

E.J. Benner Descriotion: Haddam Neck has two Copes-Vulcan PORVs both of which failed a surveillance test on 02/19/94 due to a leak in the diaphragm assembly. During the 1993 refueling outage which ended in July 1993, the licensee replaced the PORV diaphragms with a new type made of different material and of a changed shape. A lubricant was needed to help install the diaphragms due to the changed shape. The 06/30/96 A.E-10 NUREG-0933

i O

lubricant was believed to have allowed some extrusion of the diaphragm from h

between the base and the cover away from the bolt holes. The extrusion caused tears at several bolt holes and allowed the bolts to loosen over time and air to escape. An evaluation of previously reported diaphragm failures disclosed other failure mechanisms.

Oriainatina Document: Event Notification 28923.

Reaulatory Assessment: PORVs are designed to remain operable for 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> during a DBA and to be capable of for valve strokes during feed-and-bleed scenarios.

PORVs may be supplied air to open from either non-safety-related air compressors or an emergency air accumulator. The non-safety-related air compressors cannot be assumed to operate during an accident since the air compressors are not designed to operate in the harsh environment that may exist during the time when the PORVs may be required to open. Once the comprcssed air in the emergency air accumulator has been depleted, the PORVs fail closed, and core cooling capability is lost if auxiliary feedwater and main feedwater are not available for steam generator cooling. This concern is limited to those plants with high head injection pumps incapable of lifting pressurizer safety valves, for which extended inoperability of the PORVs may be highly risk significant.

The issue of pressurizer PORV relMility has been addressed in GL 90-06,

" Resolution of Generic Issue 70, ' Power-0perated Relief Valve and Block Valve Reliability,' and Generic Issue 94, ' Additional Low-Temperature Overpressure Protection for Light-Water Reactors,'" in which the staff requested licensees to include the PORVs and PORV control air system in their ASME Section XI IST

[mt Program. An IN was issued tc alert licensees to PORV failure mechanisms which may V

render PORVs degraded in significantly shorter time than common IST surveillance frequencies, thereby resulting in unexpected inoperability of the PORVs following certain DBAs.

fLqrolution: IN 95-34, " Air Actuator and Supply Air Regulator Problems in Copes-Vulun Pressurizer Power-0perated Relief Valves," dated 08/25/95.

l O mo',m lon Date: 08/25/95 GCCA-0014: DEFICIENCIES IDENTIFIED DURING ELECTRICAL DISTRIBUTION SYSTEM INSPECTIONS TAC No.: M91448

Contact:

S.S. Koenick

==

Description:==

This IN supplement represents the closeout of the electrical distribution system functional inspection. This supplement providae 4ditional information on deficiencies identified during the functional ections.

Furthermore, this supplement provides references for previousi.,

Jentified deficiencies.

Oriainatina Document: TI 2515/107, " Electrical Distribution System Functional Inspection," issued 10/19/90.

O Reaulatory Assessment: TI 2515/107 was issued in response to electrical

('j distribution system deficiencies identified by various multi-discipline inspection efforts. The TI provided instruction that allowed for a consolidated 06/30/96 A.E-11 NUREG-0933

and consistent inspection of the utilities' electrical distribution system. Any safety concerns identified during the particular inspections would be appropriately dispositioned in the respective inspection reports. The IN and the prior supplements allow for a forum to share this information with the rest of the industry.

Resolution: IN 91-29, Supplement 3, " Deficiencies Identified During Electrical Distribution System Functional Inspections," was completed in the spring of 1994 and issued on 11/22/95.

.Gomoletion Data: 11/22/95

$_CCA-0015: SEISMIC ADE0VACY OF THERMO-LAG PANELS TAC No.: M91531

Contact:

T.J. Carter

==

Description:==

A concern regarding seismic adequacy of fire barriers that use Thermo-Lag was raised by the Nuclear Information and Resource Service in a 07/21/92 petition [See Accession No. 9208280125]. The staff, based upon available information on physical properties of Thermo-Lag and analysis by a consultant of Thermal Science, Inc., concluded that the concern was not credible [See Accession No. 9302110146]. A subsequent submittal containing results of simulated seismic test and mechanical properties tests related to the use of Thermo-Lag fire barrier material at Watts Bar utilized significantly ' wer mechanical properties values compared to those used by the consultant. As a result, the NRC became concerned that the actual properties might be sufficiently different and that performance of Thermo-Lag might not be acceptable.

Oriainatina Document: A letter from TVA to the NRC on 11/11/94 [See Accession No.

9411250234].

Reaulatory Assessment: In December 1994, the staff sent followup letters to GL 92-08, "Thermo-Lag 330-1 Fire Barriers," to request additional information, including specifics on the mechanical properities. Licensees have been alerted to the concern and action has been taken to ultimately resolve the concern. This information notice clearly explains the concern and indicates how the concern will be resolved.

Resolution: IN 95-49, " Seismic Adequacy of Thermo-Lag Panels," was issued on 10/27/95 to inform licensees of concerns involving actual properties of Thermo-Lag panels.

Comoletion Date: 10/27/95 GCCA-0016: CAPABILITY OF 0FFSITE POWER DURING DESIGN BASIS EVENTS TAC No.: M91533

Contact:

T. Koshy

==

Description:==

On 01/05/95, Palo Verde reported to the NRC that, under certain offsite power system grid voltage levels, the performance of the automatic accident mitigation would be uncertain. This problem was introduced through a licensee resolution to a voltage regulation problem. The automatic loading of the 06/30/96 A.E-12 NUREG-0933

h 1

Os ECCS could be subjected to a " double sequencing" as a result of the anticipated system performance. This problem appeared to be a plant-specific concern that was discovered by the licensee in the design bases reconstitution effort. The later discussions revealed the possibility of gradual changes in switchyard voltage profile and load changes within the plant as potential causes for this scenario, j

When grid voltage is at a minimum acceptable level, the addition of safety injection loads to the bus leads to further drop in grid voltage and results in a loss of AC power. No design requirements were added into the regulation since it was not considered to be a credible event in the early evolutions of power system design. The compensatory actions that can be taken are simple in nature i

when the control room operators are sensitized to this potentially remote vulnerability. On 07/08/95, Diablo Canyon 1 & 2 reported a similar problem.

J j

Oriainatina Document: 10 CFR 50.72, Event Notification 28210 and 29168.

)

Reaulatory Assessment: When the NRC was made aware of this scenario, the staff i

evaluated the Palo Verde corrective actions. The short-term corrective action was 1

to prevent automatic loading of safety injection loads to a marginally acceptable offsite power source. The long-term solution is to appropriately evaluate the capability of the electrical system to support emergency loads and revise the setpoints for electrical bus transfer and emergency diesel generator loading.

Even though this scenario can cause unavalability of certain safety systems, the probability of such occurence is very low as it would require marginally

p acceptable grid voltage and a valid demand for safety injection. Therefore, an t

information notice was chosen as the suitable regulatory response to sensitize 4

the industry about this potential vulnerability, and to reinforce the need for l

the periodic review of electrical system capability to support accident j

mitigation.

4 j

Resolution: IN 95-37, " Inadequate Offsite Power System Voltages during Design Basis Events."

Completion Date: 09/07/95 GCCA-0017: POTENTIAL FOR LOSS OF AUTOMATIC ESF ACTUATION TAC No.: M91534

Contact:

E.N. Fields Descriotion: The Salem and Diablo Canyon licensees r1 ported a condition that could have resulted in the failure of one or both trains of the SSPS during a seismic event or a main steamline break in the turbine building (TAC No. M91479).

IN 95-10 was issued to inform other licensees of the findings by the Diablo Canyon and Salem licensees. Both licensees attempted to implement a design change that they felt would eliminate the design vulnerability. The Diablo Canyon licensee was successful in making the design change; however, the Salem licensee encountered numerous problems in attempting the design change.

Supple:nent 1 to IN 95-10'provided details on the problems the Salem licensee exps. lenced, including inadvertently de-energizing the source range high voltage block signal and the problems encountered in attempting to remove the SSPS logic matrix power supplies.

06/30/96 A.E-13 NUREG-0933

I Supplement 2 to IN 95-10 was issued on 08/11/95 primarily to communicate the Salem licensee's corrective actions with respect to the surveillance and 4

maintenance practices for the SSPS. The licensee implemented these corrective actions in an effort to improve the reliability of the SSPS. It appeared worthwhile to apprise the industry of these actions.

Oriainatina Document: 10 CFR 50.72 Event Notification 28318.

Reaulatory Assessment: Shortly after the NRC was made aware of the SSPS design deficiency and the problems encountered by the Salem licensee in attempting corrective actions, affected licensees were apprised of their potential susceptibility to the problem. Licensees were notified through direct contact by NRC project managers, through H notification, and through IN 95-10 (and Supplement 1). Therefore, there is no immediate safety concern with regard to the timing of the issuance of this IN Supplement 2 and the overall timing of the staff's efforts in addressing the concerns that have surfaced from this event has not been inappropriate.

Resolution: IN 95-10, Supplement 2, " Potential for Loss of Automatic Engineered Safety Features Actuation," dated 08/11/95, and Memorandum from B. Boger to W.

Russell dated 04/11/95.

Completion Date: 08/16/95 GCCA-0018: POTENTIAL FOR MOV FAILURE - STEMIILOIICTION PIPE CHANGES TAC No.: H91624

Contact:

T.A. Greene Descriptie: A RHR MOV at Cooper failed to closed on demand while the RHR trains

{

were being switched. Licensee reviews showed that the stem protector pipe on the valve actuator had threaded into the M0V housing and interfered with the stem nut rotation. The motor was damaged during an attempted opening stroke. The stem protector pipe had previously been replaced by the licensee. The replacement stem protector pipe was manufacture by the licensee. However, it was not constructed to the same tolerance of the original. Specifically, the length of threaded portion of the protector pipe was too long when manufactured. The extended threads allowed the pipe to be threaded to the point where it interfered with the stem nut locknut. Also, actions (such as staking) were not taken to prevent threading of the stem protection pipe into the valve actuator housing.

Oriainatina Document: Morning Report 4-95-0014.

Reaulatory Assessment: A stem protector pipe may be attached to an M0V through the housing cover to prevent debris from entering the stem / actuator interface area. To keep the stem protector pipe from interfering with actuator operation (specifically rotation of the stem nut and its locknut), the threads on the pipe may be restricted to a certain length. Another option is to stake the threads on the stem protector pipe at a specific location. If neither of these precautions is taken, the stem protector pipe may thread sufficiently into the actuator housing to interfere with the rotation of the stem nut locknut. Additional torque may be required to operate the valve, which may cause the torque switch to trip prematurely, motor thermal overload devices to activate, or the motor to be damage on high torque demand.

06/30/96 A.E-14 NUREG-0933

i Resolution: At Cooper, the licensee long-term solution was to stake the extended V) threads in all of the stem protector pipes of safety-related MOVs. IN 95-31, t

4

" Motor-0perated Valve Failure Caused by Stem Protector Pipe Interference," was.

issued on 08/09/95 to address the generic concern.

4 Comoletion Date: 08/09/95 GCCA-0019: UNANTICIPATED AND UNAUTHORIZED MOVEMENT OF FUEL TAC No.: M91642

Contact:

C.V. Hodge Descriotion: IN 94-13 implied a broader applicability of training requirements j

of 10 CFR 50.120 to contract personnel than is the case. In October 1994, DRCH/NRR proposed a supplement to IN 94-13 to clarify confusion of the original i

notice concerning application of 10 CFR 50.120. The descriptive language in the i

original notice implied a broader applicability to contract personnel than is the case. In addition, in March 1995, there were contractor mistakes at Hatch that resulted in a dropped core shroud bolt that punctured the spent fuel pool liner, dropped stellite bearings on the transfer canal floor that proddced an uncontrolled high radiation area, and placement of excessively contaminated items j

outside of radiologically controlled areas.

j Oriainatina Document: IN 94-13, " Unanticipated and Unintended Movement of Fuel Assemblies and Other Components Due to Improper Operation of Refueling l

Equipment."

Reaulatory Assessment: On 03/31/95, DRCH concurred in authorizing publication of a supplement to IN 94-13 to address the original contract personnel applicability j

question and the Hatch events as additional examples of problems associated with the control of contract personnel activities and inadequate oversight of j

refueling operations.

Resolution: The supplement was issued on 11/28/95.

I Comoletion Date: 11/28/95 GCCA-0020: FRAUDULENT COMMERCIAL GRADE CERTIFICATE OF COMPLIANCE TAC No.: M91643

Contact:

C.V. Hodge

==

Description:==

In September 1989, Southern Testing Services contracted with Relay Specialties, an authorized Agastat relay distributor, to provide: (1) 32 Agastat commercial-grade Model 7022AC relays; (2) Agastat relay sockets, P/N 700137; and (3) a " certificate of conformance for [the relays) from Agastat." In April 1991, during an NRC inspection at the Southern Testing Services facility, an NRC inspector became suspicious of an Amerace certificate of compliance because it did not appear to have the correct Amerace facility addrecs. During subsequent discussions between NRC and Amerace personnel, the NRC inspector found that the certificate of compliance had not been issued by Amerace, nor was the signature on the certificate of compliance that of an Amerace employee. Amerace stated that

(

the correct Amerace facility address on a certificate of compliance for that particular time period was 530 W. Mt. Pleasant Avenue, Livingston, New Jersey.

06/30/96 A.E-15 NUREG-0933

In contrast, the address shown on the attached certificate of compliance is 190 Lincoln Highway, Edison, New Jersey. The NRC inspector examined one of the Agastat relays that Southern Testing Services had used as a test specimen during the commercial-grade component testing and noted that the label affixed appeared to be a label used by the Control Component Supply company for relays that had been field-modified. Control Component Supply was an authorized distributor of Amerace relays and related components.

Oriainatina Document: Special Inspection Report 99900289/91-01.

Reaulatory Assessment: Agastat relays are used in numerous safety-related applications in nuclear power plants. Because Amerace supplied these relays as commercial-grade components, the whole nuclear power industry could theoretically procure them and dedicate them to safety-related service.

Resolution: RVIB drafted an IN. Due to the nature of the concern, concurrence in the IN was needed from OGC who asked questions regarding the development of additional events since the inspection. The IN was cancelled based on the age of the issue and the lack of any additional occurrences since the inspection.

Comoletion Date: 07/20/95 GCCA-0021: CHATTER OF ITT BARTON 288A AND 289A DIFFERENTIAL PRESSURE TRANSMITTERS TAC No.: M91645

Contact:

C.V. Hodge

==

Description:==

ITT Barton is a supplier of basic components used in safety-related applications in nuclear power plants. The company has supplied Models 288A and 289A differential pressure (dp) indicating switches with certification to IEEE-323/344 qualifications. These include no switch contact chatter during seismic loadings of up to 129 for monitored dp different from setpoint by at least 10%

full scale. In 1994, unrelated testing revealed potential concerns regarding contact chatter. As a result, additional testing of qualified configurations of these products was performed at Wyle Laboratories in Norco, CA, during the first week of 1995.

Oriainatina Document: Part 21 Report from ITT Barton on 10/05/94 [ Accession No.

9411010149].

Reaulatory Assessment: The safety significance and generic implications depend on how the transmitter is used. The contact chatter at 129 has reasonable probability of occurrence if the instrument is located at a high elevation and the plant is built on soil rather than rock. An SSE is usually characterized by a smaller acceleration (0.2g on the east coast, 0.5g on the west coast), but the mechanical parameters displacement, velocity, and acceleration may be amplified for various frequencies of oscillation. (Consider the mechanical model of a mass coupled through a damped spring to an oscillating base.) As explained in the Part 21 report dated 01/19/95,' int by 10% fullthe chatter does not occur when the monitored d different from the setpo scale. The report also notes that

" testing has again verified that these products are capable of surviving earthquake loadings up to 12g with no degradation in structural or pressure boundary integrity and no degradation in functionality with regard to pointer 06/30/96 A.E-16 NUREG-0933

i O

this report because of QA aspects of safe operation of nuclear power plants. This accuracy, switch setpoint change and switch deadband." The NRC is concerned with vendor has evidently provided production models of this instrument, samples of which did not perform well on qualification tests. It follows that a production i

control problem may exist. No event reports have been received similar to this Part 21 report. This report relates to GI A-46, " Seismic Qualification of Equipment in Operating Plants," and GI 114, " Seismic-Induced Relay Chatter,"

identified in March 1985. In the event of an earthquake of sufficient magnitude (g level) that causes relay chatter, combined with simultaneous loss of offsite power and potential misalignment of equipment, instrumentation, and circuit breakers, core damage may result, absent corrective operator action.

Resolution: In a subsequent Part 21 report dated 02/15/95, the vendor supplied a list of customers. RVIB reviewed the engineering report and conducted a vendor inspection. The following excerpt from Inspection Report 99900113/95-01, pp. 5-6, Sec 3.6, demonstrates that this concern is closed because corrective vendor action was sufficient.

"Barton began an engineering evaluation of mild environment equipment qualification in late 1994 as a result of a licensee group audit. During review of 1980 and 1986 seismic test reports, Barton determined that switch chatter may have occurred that was not detected. The specific instrumentation used to monitor contact chatter was not identified in the test reports, but was suspected of being an incandescent lamp. Barton conducted additional seismic tests early in January 1995, using instrumentation capable of measuring contact chatter as rapid as two milliseconds. The new testing showed that higher g levels, and setpoints very close to actual parameter values, produced the most chatter; there was no chatter at 4g. Barton provided all affected cgstomers with a table showing the duration of contact chatter as a function of g level and the proximity of the trip setpoint to the actual differential pressure value. The NRC inspector discussed this with Barton. The review of data from earlier seismic tests, and conducting additional tests, revealed a possible concern that had gone unnoticed for several years. The inspector considered Barton's activities including notifications to be acceptable, and no further action is required."

No NRC generic response, such as an IN, is needed. Two previous ins are related to this issue: IN 85-02, " Improper Installation and Testing of Differential Pressure Transmitters," and IN 86-65, " Malfunctions of 11T Barton Model 580 Series Switches During Requalification Testing."

Comoletion Date: 07/20/95 GCCA-0022: FREQUENCY OF USE OF AIR-OPERATED GATE VALVES TAC No.: M91746

Contact:

C.V. Hodge Descriotion: Air-operated gate valves, utilizing Hillar actuators, in the room cooler water supply lines failed to open (safety function) during surveillance.

The problem seems to be related to the frequency of operation. Hope Creek has 32 valves of this type.

Oriainatina Documents: 10 CFR Part 50.72, Event Notification 28014, dated 11/10/94; LER 50-354/94-17, dated 12/08/94.

06/30/96 A.E-17 NUREG-0933

Reaulatory Assessment: The LER explains that Hope Creek systems engineering determined a common cause (packing design). The probable root cause is packing configuration combined with a long interval between valve stroking. These valves were subject to a design change from 9 rings of Crane packing to 9 rings of Chesterton Graphfoil packing. The industry standard configuration is 4 or 5 rings of graphite packing with carbon busing. The standard packing gland torque calculations are based on 5 rinas. By calculation, reduction from 9 to 4 or 5 rings reduces the dynamic packing load up to 50% and reduces the total force needed by the actuator. A contributing cause may involve use of the disk friction coefficient from the original sizing calculations. Such usage is generally non-conservative, according to findings from GL 89-10, " Safety-Related Motor-Operated Valve Testing and Surveillance." Industry experience shows that long stationary times foster packing sticking. Thus, static packing loads increase substantially and the total load may exceed the capability of the actuator. Also, a higher stem finish is needed for graphite packing. Safety-related stuck air valve events have been reported from Peach Bottom.

Resolution: The licensee reported that the problem related to the packing changeout was the failure to adjust the air regulator to accommodate the lower packing friction. There was some evidence that the force of the actuator drove the valve disk into the seat and then the spring force could not dislodge it within the required stroke time. Further discussions with the licensee indicated that a contributing factor, which may have been the overriding reason for failure, was that the licensee failed to install mechanical stops on the valve.

These stops prevent the valve actuator form driving the valve beyond its closed position. While valve packing issues are generic to the industry, the main cause appears to be related to the failure to install the stops which is not a generic issue. Therefore, an IN was not deemed necessary.

Comoletion Date: 08/31/95 GCCA-0023: PRESSURE LOCKING AND THERMAL BINDING OF GATE VALVES TAC No.: d91781

Contact:

J.W. Shapaker Descriotion: In GL 89-10 (06/28/89), the NRC staff asked addressees to provide additional assurance of the capability of safety-related M0Vs and certain other M0Vs in safety-related systems to perform their s'afety-related functions. In Supplement 6 to GL 89-10 (03/08/94), the NRC staff provided guidance on an acceptable approach for addressing pressure locking and thermal binding of MOVs, but did not request specific actions. During inspections of GL 89-10 programs, the NRC staff found the actions taken by licensees to address pressure locking and thermal binding of MOVs to be varied. In view of these inspection results, and the fact that most addressees were nearing completion of their GL 89-10 programs, the NRC staff determined that issuance of a subsequent GL was warranted to request that addressees perform, or confirm that they previously performed:

(1) evaluations of operational configurations of safety-related, power-operated (including motor,

air,

and hydraulically-operated) gate valves for susceptibility to pressure locking and thermal binding; and (2) further analyses and any needed corrective' actions to ensure that safety-related, power-operated gate valves that are susceptible to pressure locking or thermal binding are capable of performing the safety functions within the current licensing bases of the facility.

06/30/96 A.E-18 NUREG-0933

1 i

O Oriainatina Documents: GL 89-10, " Safety-Related Motor-0perated Valve Testing and

)

V Surveillance," dated 06/28/89; GL 89-10, Supplement 6, " Safety-Related Motor-Operated Valve Testing and Surveillance," dated 03/08/94.

Reaulatory Assessment: 10 CFR 50 (Appendix A, Criteria 1 and 4) and plant I

licensing safety analyses require and/or commit that the addressees design and test safety-related components and systems to provide adequate assurance that i

these systems can perform their safety functions. Other individual criteria in Appendix A to 10 CFR Part 50 apply to specific systems. In accordance with these regulations and licensing commitments and under the additional provisions of 10 CFR 50 (Appendix B, Criterion XVI), licensees are expected to take actions to ensure that safety-related, power-operated gate valves susceptible to pressure locking or thermal binding are capable of performing their required safety functions. Supplement 6 to GL 89-10 alerted licensees to the problems with pressure locking and thermal binding in MOVs and described an acceptable approach for addressing these phenomena for MOVs, but did not request any specific actions or response from licensees.

1 The actions requested in GL 95-07 are considered compliance backfits, under the provisions of 10 CFR 50.109 and existing NRC procedures, to ensure that safety-related, power-operated gate valves that are susceptible to pressure locking or thermal binding are capable of performing their intended safety functions. In accordance with the provisions of 10 CFR 50.109 regarding compliance backfits, a full backfit analysis was not performed. However, the staff performed a documented evaluation which stated the objectives of and reasons for the requested actions and the basis for invoking the compliance exception. This m[V i

evaluation is available in the NRC PDR [ Accession No. 9508240264].

Resolutio11: Issuance of GL 95-07, " Pressure Locking and Thermal Binding of Safety-Related Power-0perated Gate Valves," dated 08/17/95.

[omoletion Date: 08/17/95 GCCA-0024: DEGRADED DECAY HEAT REMOVAL CAPABILITY VIA NATURAL CIRCULATION TAC No.: M91805

Contact:

J.R. Tappert DescriptiQn: The ability of steam generators (SGs) to remove decay heat via sub-cooled natural circulation may be degraded under certain plant conditions; specifically, when the RCS is vented. This limitation is not explicit in TS and some licensees have relied on SGs as one of their two redundant sources of decay heat removal when they may not have been able to fully perform this function.

Oriainatina Document: Memorandum from E. Merschoff to B. Grimes dated 03/09/95.

Reaulatory Assessment: TS generally require two methods of decay heat removal in Mode 5 with loops filled. When this is the case, they generally go on to indicate that this requirement can be satisfied by two loops of RHR or one loop of RHR and a minimum water level in the SGs. Decay heat can be removed either through the RHR system or through the SGs by natural circulation af ter the RCPs are secured.

/G The heat removal mechanism with RHR is through forced circulation through the RHR Q

heat exchanger. Heat removal with natural circulation of reactor coolant through the SGs occurs because of the differential pressure created between the heated 06/30/96 A.E-19 NUREG-0933

water in the reactor core and the cooler water in the SG tubes. This differential pressure is created through temperature differences that in turn create fluid density differences between these two locations.

During natural circulation, the SG secondary side water boils and steams off through the atmospheric relief valves or other openings that may exist during shutdown conditions. The minimum temperature at which boiling will begin in the SG is 100 C [212 F]. A minimum temperature differential of 28 C [50 F] between the RCS and the SG secondary water is routinely used for evaluating conditions that would ensure sufficient natural circulation flow to prevent boiling in the core. The heat transfer rate across the SG tubes is less for lower RCS-to-SG secondary temperature differentials but still may be adequate to promote sufficient natural circulation and prevent core boiling. Adding the differential l

temperature of 28 C [50 F] to 100 C [212 F] results in a minimum RCS temperature of 128*C [262 F] to maintain sufficient natural circulation flow.

The lowest pressure point in the RCS, at the top of the SG tubes, should therefore be maintained above the saturation pressure for 128 C [262aF]. If the RCS pressure at the top of the SG tubes is allowed to fall below the primary fluid saturation temperature, flashing and steam voiding may occur, interrupting or degrading the natural circulation flow path. Additionally, when system pressure is dropped with elevated water temperatures, gases may come out of solution.

When the RCS is being depressurized and cooled down, the RCPs are stopped, the RCS is depressurized and vented, and level is decreased in preparation for Mode 6 (refueling) entry. In Mode 6, both RHR trains must be operable. During the transition from Mode 5, with no RCPs running, to Mode 6, plant conditions may exist that are not adequate to support natural circulation. The second train of RHR may need to be operable before proceeding with plant cooldown and depressurization to provide a second method for RCS cooling.

Some licensees may not clearly understand that plant conditions may degrade the SGs ability to remove decay heat by natural circulation. An IN was written to alert them to these conditions. Due to the fact that at least one train of RHR is always required, an IN was an adequate response to this issue.

Resolution: On 08/28/95, IN 95-35, " Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation," was issued to alert licensees to relevant plant operating experiences.

Comoletion Date: 08/25/95 GCCA-0025: FALSIFICATION OF ASNT CERTIFICATE BY AMERICAN POWER SERVICES TAC No.: M91950

Contact:

T.A. Greene

==

Description:==

American Power Service (APS) of Georgetown, Massachusetts, deliberately gave falsified American Society for Nondestructive Testing (ASNT) certificates to an NRC licensee in connection with the procurement of commercial-grade services. Northeast Nuclear Energy Company (NNEco) had contracted with APS for commercial-grade services in supplying rebabbitted bearings for the Millstone-2 reactor building closed cooling water system pump. Since APS did not have a NNEco approved QA program, the re-babbitted bearings were to be dedicated and accepted under NNEco's QA program. The dedication process included ultrasonic 06/30/96 A.E-20 NUREG-0933

. =

4 A

i examination which was to be performed by NNECo as part of the receiving inspection.

s During NNEco source inspection of APS, the APS president informed the NNECo Quality Assessment Services (QAS) inspector that APS had developed in-house UT capability and gave the inspector copies of what appeared to be ASNT certificates j

identifying three of the company's employees as UT inspectors or examiners. Upon observing that the certificates had identical serial numbers, the NNECo QAS inspector contacted ASNT and learned that ASNT had rot issued the certificates i

in question.

Oriainatina Document: Letter from NNEco to NRC dated 03/04/94 (Accession No.

9403150377).

i Reaulatory AsjigjigJn_ed: Since APS was performing a commercial-grade service and since the acceptance of the bearings was based on UT performed by NNEco, the integrity of the bearings was not compromised and NNECo was able to justify their j

use in the intended application.

Resolution: The NRC Office of Investigations determined that the president of APS j

deliberately and improperly falsified ASNT certificates and gave them to NNECo.

The NRC informed APS of its findings by letter dated 02/14/95 (Accession No.

i 9502160007). The generic concern was resolved by the issuance of IN 95-45, i

"American Power Service Falsification of American Society for Nondestructive j

Testing (ASNT) Certificates."

f h)

Comoletion Date: 10/4/95

V 4

GCCA-0026: ADEQUACY OF EMERGENCY AND ESSENTIAL LIGHTING i

TAC No : M91952 fan 13d: N.K. Hunemuller

==

Description:==

The objective of the NRC requirements and guidelines for emergency

)

lighting is to ensure that in the event of a fire, plant personnel can access and operate equipment and components that must be manually operated to effect safe plant shutdown. Since IN 90-69 was issued on 10/31/90, there have been several reported events of deficient emergency lighting, involving Vermont Yankee, Cooper, Indian Point 3, Washington Nuclear 2, and Diablo Canyon. NRC inspections 4

i have also uncovered emergency lighting deficiencies.

Oriainatina Document: Event Notification 28025, dated 11/12/94.

i Reaulatory Assessment: During a meeting on 04/04/95, the Events Assessment and Generic Issues Panel approved the preparation of an IN on potential problems with post-fire emergency lighting. IN 95-36, " Potential Problems with Post-Fire 1

Emergency Lighting," was issued on 08/29/95. This notice alerts addressees to potential problems regarding emergency lighting for plant areas needed for 1

operation of post-fire safe shutdown equipment and in access and egress routes thereto.

J p Resolution: IN 95-36, " Potential Problems with Post-fire Emergency Lighting,"

dated 08/29/95, was issued to alert licensees to relevant plant operating l\\

experiences but does not require any licensee action.

. 'j 06/30/96 A.E-21 NUREG-0933 a

Comoletion Date: 08/29/95 GCCA-0027: ADDRESS CONCERNS REGARDING ASME CODE TAC No.: M91965

Contact:

R.A. Benedict

==

Description:==

The staff was concerned about: (1) the use of engineering judgment regarding the ASME Code; (2) the NRC's position on the regulatory status of ASME Code interpretations prepared by ASME; and (3) inaccuracies in GL 90-05 and IN 93-21.

Oriainatina Document: None.

Reaulatory Assessment: (1) Engineering judgment cannot replace NRC requirements, whether or not these requirements contradict the ASME Code. The licensee must request relief, not use engineering judgment, when the provisions of 10 CFR 50.55a are considered. (2) NRC endorsement in 10 CFR 50.55a is limited only to those editions / addenda of the ASME Code that are specifically identified and approved. (3) The staff discussed the differences between weld overlay and weld buildup and how these may be applied to Code requirements (IN 93-21). The staff also noted that pressure boundary leaks on safety-related systems always require relief when a Code repair is not performed (GL 90-05).

Resolution: Issuance of a letter from B.W. Sheron to R.F. Reedy on 07/24/95.

Comoletion Date: 07/24/95 GCCA-0028: CIRCUMFERENTIAL CRACKING OF STEAM GENERATOR TUBES TAC No.: M92004

Contact:

E.J. Benner

==

Description:==

Since the issuance of GL 95-03, "Circnferential Cracking of Steam Generator Tubes," additional information pertaining to the tubes removed from Maine Yankee for destructive analysis has become available. In addition, the wrong title was given to NUREG-0844 in GL 95-03 as, " Voltage-Based Interim Plugging Criteria for Steam Generator Tubes." The correct title is, "NRC Integrated Program for the Resolution of ' Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity." This TAC was initiated to develop an IN to correct the error in the GL and to provide additional information to the licensees.

Oriainatina Document: GL 95-03, "Circumferential Cracking of Steam Generator Tubes," dated 04/28/95.

ELquiatory Assessment: The staff issued GL 95-03 to obtain information necessary to assess compliance with requirements regarding steam generator tube integrity in light of the inspection findings at Maine Yankee. In GL 95-03, the staff requested that utilities: (1) evaluate recent operating experience with respect to the detection and sizing of circumferential indications; (2) develop a safety assessment justifying continued operation until the next scheduled steam generator tube inspections are performed; and (3) develop plans for the next inspections of steam generator tubes as they pertain to the detection of 06/30/96 A.E-22 NUREG-0933

I circumferential cracking. The error and additional information described above Q

under " Description" do not change the basis for the actions required in the GL.

Therefore, issuance of an IN is an appropriate regulatory action to correct the error and provide licensees with the additional information.

Resolution: IN 95-40, " Supplemental Information to Generic Letter 95-03,

'Circumferential Cracking of Steam Generator Tubes,'" was issued on 09/20/95.

Comoletion Data: 09/20/95 GCCA-0029: REACTOR COOLANT PUMP TURNING VANE BOLT LOCKING DEVICE FAILURE TAC %.: M92027

Contact:

E.N. Fields

==

Description:==

On 06/03/94, the licensee for Seabrook conducted an underwater examination of reactor vessel internals. Foreign material was found on the reactor vessel internals lower core plate. In a subsequent video inspection, a bolt was found on the bottom of the reactor vessel and two bolt-locking devices were found on the lower core plate. One locking device was intact and the other was deformed.and had portions missing. The licensee identified the bolt and locking devices as a cap screw and locking cups that are used in the RCPs to attach and secure the turning vane diffuser to the thermal barrier flange.

The degradation of the locking device and the release of the bolt was evaluated f-by the licensee with assistance from W. The root cause of the release of the turning vane cap screw and locking cups was attributed to the original design not adequately considering the affects of flow-induced vibration on the locking cup and the turning vane cap screw. The licensee postulated that flow-induced vibration caused the locking cups to erode and release from the turning vane. The cap screw subsequently backed out as a result of the loss of the pre-load torque and the effects of vibration and gravity.

Oriainatina Document: LER No. 94-010-01.

Reaulatory Assessment: From a safety perspective, the failure of a locking cup and the resultant release of a turning vane cap screw, in the worse case, could result in fuel damage and/or subsequent fuel failure. Licensees would likely be alerted to these failures by the loose parts monitoring system. However, the point at which loose parts noise activity would force a reactor shutdown is largely site-specific. If fuel damage resulted from the impact of loose parts, the licensee would be alerted by increased RCS activity, either by radiation monitors or required RCS sampling. Maximum allowable RCS activity is controlled by TS. Any increase beyond TS limits would force a reactor shutdown to evaluate the source of the activity.

Resolution: IN 95-43, " Failure of the Bolt-Locking Device on the Reactor Coolant Pump Turning Vane," was issued on 09/28/95.

Comoletion Date: 09/28/95 06/30/96 A.E-23 NUREG-0933

GCCA-0030: MAIN STEAM ISOLATION VALVE FAILURE DUE TO PILOT VALVE MALFUNCTION TAC No.: M92028

Contact:

D.L. Skeen

==

Description:==

Sticking solenoid pilot valves (Automatic Switch Company [ASC0]

model NP8323) caused two MSIVs to fail to close at LaSalle-2; additionally, one MSIV at LaSalle-1 closed after a 15-second delay. These events occurred in February and June 1995, respectively. The root cause was determined to be contamination of internal piece parts (the core assembly and plug nut) that caused them to stick together. The contamination is believed to be a combination of lubricant (Nyogel 775A) used by the manufacturer during assembly of the NP8323 SOVs and thread sealant used when connecting the air lines to the pneumatic actuator, of which the NP8323 valve is a part.

The manufacturer discontinued production of the model NP8323 S0V in 1990 due to problems with foreign material entering the i alves and causing either inhibited movement or degradation of the ethylene propylene elastomers. The manufacturer believed that the inhibited movement was the result of lubricant or thread sealant from the MSIV pneumatic actuator assembly and the degradation of the elastomers was the result of ester oils used in the air compressors of the instrument air systems at nuclear power plants.

Even though ASCO believed that field conditions rather than design defects were responsible for the problems seen in the NP8323 valves through 1990, they notified customers at that time that the NP8323 S0Vs were being discontinued and should be replaced in a timely manner. Some customers replaced the SOVs, but other customers who had not experienced any problems ordered more of the NP8323 valves before production was halted. Since it was not clear how many licensees may still be using these SOVs, an IN was issued by the NRC to alert licensees to the recent problems experienced by LaSalle.

Qriginatina Documents: NRC Region III Morning Report 3-95-0019, dated 02/22/95; 10 CFR 50.72, Event Notification 28926, dated 06/11/95.

Reaulatory Assessment: The MSIVs function to limit the release of radioactive materials to the environment or limit reactor coolant inventory loss given a steam line break. There are two MSIVs in series for each main steam line either one of which is capable of isolating the line if needed. The scenario of concern would be a steam line break in a line where both MSIVs failed to close because of sticking pilot valves.

Resolution: IN 95-53, " Failures of Main Steam Isolation Valves as a Result of Sticking Solenoid Pilot Valves," was issued on 12/01/95.

Comoletion Date: 12/01/95 GCCA-0031: POTENTIAL CABLE DAMAGE FROM EXCESS SIDE WALL PRESSURE TAC No.: M92215

Contact:

J.L. Birmingham

==

Description:==

This proposed IN discussed the concern that some licensees had not considered cable sidewall bearing pressure when calculating allowable pull tension. The proposed IN also discussed instances that occurred during the past 06/30/96 A.E-24 NUREG-0933

five years in which cable sidewall bearing pressure requirements had been

-\\

exceeded or had not been documented in procedures used for cable pulls. However, 4

these instances did not result in identified cable damage.

Oriainatina Document: The IN Authorization Form (from R.L. Spessard to A.E.

1 Chaffee) signed on 3/27/95 proposed an IN on failure to consider cable sidewall bearing pressure during cable pulls.

Reaulatory Assessment: During review of the proposed notice, a thorough search of NRC and industry data sources for similar occurrences was made. This search did not find any additional recent examples of failure to consider cable sidewall bearing pressure. The NRC had addressed an extensive failure to consider cable sidewall bearing pressure that occurred at Watts Bar-2. As a result of that occurrence, the NRC issued IN 92-01, " Cable Damage Caused by Inadequate Cable Installation Procedures and Controls." Because the concern does not appear to be a current problem and because the concern was previously addressed by the NRC in a generic communication, issuance of the proposed notice is not necessary at this time.

Resolution: The need for the proposed IN was discussed at a meeting of the Events Assessment Panel on 08/01/95 and the proposed notice was determined not to be necessary and was cancelled. The determination was based on the concern being i

primarily plant-specific.

l Completion DaAq: 08/01/95 O

GCCA-0032: EVALUATE MISSILES FROM MIRROR INSULATION DURING HIGH ENERGY PIPE 4

BREAKS TAC No.: M92216

Contact:

J.L. Birmingham

)

==

Description:==

During the NRC staff's evaluation of actions to prevent clogging of f

suppression pool strainers, a concern was identified that, in the event of a l

LOCA, reflective metallic insulation (RMI) debris may form missiles and damage l

safety system components such as control or power cabling.

Oriainatina Document: Chapter 5 of the Draft Report, " Knowledge Base for Emergency Core Cooling System Recirculation Reliability," being prepared for the j

Organization for Economic Cooperation and Development / Nuclear Energy Agency.

Reaulatory Assessment: Initial assessment of this concern determined that the metallic foil in mirror insulation could potentially damage control or power cabling during a LOCA and affect the operability of safety systems inside containment. However, the concern was mitigated because redundant safety systems are typically isolated by distance or missile shielding and because critical components for mitigating an accident, such as ECCS pumps, are not located inside the drywell. Further, cables inside the drywell are generally in conduits or cable trays which are of a durable construction that would make it difficult for the debris to damage cabling.

O After the staff's initial assessment, two tests have been conducted which appear to substantiate the assessment. The first test was sponsored by the NRC and was conducted at the Siemens facility in Karlstein, Germany. The second test was 06/30/96 A.E-25 NUREG-0933

performed by TRANSCO Products Inc., an RMI vendor. The results of both tests showed that little significant damage was done to the cable jacket material by the metallic foil in the RMI and that the metallic foil did not damage the cable itself.

In the first test, some of the cabling was severed. This is believed to have been caused by large chunks of the RMI panels impacting on the cables. The staff noted that the cabling was hung in a manner atypical of domestic BWRs in that the cabling was not installed in cable trays or conduits and was tightly hung with little slack in the cable which did not allow the cable to give on impact with the larger RMI debris. The larger pieces of RMI debris were readily stopped by intervening structures and were found to be wrapped around them. The staff has concluded that cable trays and conduits, as well as containment structures and piping, are likely to provide additional protection from large RMI debris.

Resolution: Based on evaluation of the test results, the staff recommended that no further action be taken on this issue at this time. In addition to the low safety significance, the staff found that the issue is very plant-specific and break location-specific and is therefore best addressed by the individual licensees. The staff intends to include a discussion on this issue in the final bulletin addressing the strainer issue so that licensees who have or in%nd to install RMI will be advised of the potential problem and will account for it in their final resolution. The staff documented this resolution in a memorandum from R.B. Elliott to C. Berlinger on 11/07/95. The memorandum has additional details on the safety assessment and testing.

.Comoletion Date: 11/07/95 GCCA-0033: RESULTS OF RECENT NRC SPONSORED FLAME SPREAD AND FIRE ENDURANCE TESTING TAC No.: M92406

Contact:

T.J. Carter

==

Description:==

Flame spread tests involving Thermo-Lag 330-1 fire barrier panels were conducted 01/12/95. The results have not been provided to licensees for determination of applicability to their facilities. Previous concerns had been raised regarding performance of Thermo-Lag as a fire barrier material.

Oriainatina Document: None.

Reaulatory Assessment: Licensees have taken compensatory measures as a result of concern about the Thermo-Lag performance. Therefore, there is time to conduct tests and provide licensees with information as it becomes available.

Resolution: IN 95-32, "Thermo-Lag 330-1 Flame Spread Test Results," was issued on 08/10/95 to provide results of the flame spread tests.

Completion Date: 08/10/95 0

06/30/96 A.E-26 NUREG-0933

~_

i(

COMMISSION DECISION ON THE RESOLUTION OF GENERIC ISSUE 23. " REACTOR GCCA-0034:

COOLANT PUMP SEAL FAILURE" TAC No.: M92408

Contact:

T. Koshy

==

Description:==

On 04/19/91, the NRC published Federal Register Notica % FR 16130 requesting comments on the then-current understandings, findings, and potential i

1 recommendations regarding GI-23, together with a draft Regulatory Guide, DG-1008,

" Reactor Coolant Pump Seals." On 05/02/91, NRC issued GL 91-07, "GI-23, ' Reactor Coolant Pump Seal Failures' and Its Possible Effect on Station Blackout," which stated that preliminary results of NRC studies suggested that RCP seal leak rates could be substantially higher than those assumed in the coping analyses for implementation of the station blackout (SB0) issue. The GL reminded licensees 4

that higher seal leak rates could affect licensee analyses and actions addressing

{

conformance to the SB0 rule.

Staff studies and analyses concerning RCP seal leakage are documented in 4

Appendices A and B to NUREG/CR-5167, " Cost / Benefit Analysis for Generic Issue 23:

Reactor Coolant Pump Seal Failures," April 1991, which contains the NRC model for RCP seal failure. The report identifies several modes of RCP seal leakage which may be in excess of that assumed in licensee coping analyses for implementing the requirements of 10 CFR 50.63, the SB0 rule.

The Commission considered the proposed rulemaking as a method to resolve GI-23.

In SECY-94-225, dated 08/26/94, a draft rule was proposed for public comment that

,s would resolve GI-23. On 03/31/95, the Commission voted against publication of the

(

proposed rule that would have resolved GI-23. The Commission concluded that the proposed rule did not provide sufficient gain in safety to justify its issuance.

The Commission was also concerned that inaccuracies in the NRC seal leakage evaluation model may exist.

Oriainatina Document: SRM dated 03/31/95.

Reaulatory Assessment: The NRC staff had conducted extensive study on the vulnerability of RCP seal failures and considered issuance of generic requirements to address this issue. Since this issue was identified as a GI, the industry has addressed the problem. Improvements in RCP seal designs and the exchange of coping analyses information, etc., are the outcome of these efforts.

Even though a new rule was not issued, the staff efforts so far has resulted in increased sensitivity to this issue and has resulted in a documented coping analysis through the SB0 rule at every station. The staff continues to evaluate if any further action is needed for the disposition of this issue.

Resolution: IN 95-42, " Commission Decision on the Resolution of Generic Issue 23,

' Reactor Coolant Pump Seal Failure,'" was issued on 09/22/95.

Comoletion Date: 09/30/95 0

06/30/96 A.E-27 NUREG-0933

GCCA-0035: VOLTAGE-BASED INTERIM REPAIR CRITERIA FOR STEAM GENERATOP TUBES TAC No.: M92578

Contact:

J.W. Shapaker Descriotion: PWR facility licensees may, on a voluntary basis, request a license amendment to the plant TS to implement alternate SG tube repair criteria applicable specifically to outside diameter stress corrosion cracking (00 SCC) at the tube-to-tube support plate intersections in H-designed SGs having drilled-hole tube support plates (TSP) and Alloy 600 tubing. Guidance is offered on implementing the alternate (voltage-based) repair criteria. By approving the use of specific vcitage-based repair criteria as an acceptable measure for dealing with ODSCC tube degradation, it is intended to provide relief while maintaining an acceptable level of safety for licensees having SGs experiencing this particular degradation mechanism until a longer-term resolution to the issue of SG degradation is pursued through rulemaking.

Oriainatina Document: None.

Reaulatory Assessment: Design of the RCPB for purposes of structural and leakage i

integrity is a requirement under 10 CFR 50, Appendix A, GDC for Nuclear Power Plants. Specific requirements governing the maintenance of SG tube integrity are in plant TS and in Section XI of the ASME Boiler and Pressure Vessel Code (ASME Code). These include requirements for periodic ISI of the tubing, flaw acceptance criteria (i.e., repair limits for plugging or sleeving), and primary-to-secondary leakage limits. These requirements, coupled with the broad scope of plant operational and maintenance programs, have formed the basis for ensuring adequate SG tube integrity.

The traditional strategy for achieving the objectives of the GDC regarding SG tube integrity has been to establish a minimum wall thickness requirement in 4

accordance with the structural criteria of RG 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes." A further assumption of uniform thinning results in the development of a repair limit (40% depth-based repair limit is typically incorporated 1-to plant TS) that is conservative for all flaw types.

The NRC has approved for use, on a voluntary basis, alternate (voltage-based) repair criteria that are only applicable to predominantly axially-oriented ODSCC indications located at the tube-to-TSP intersections in H-designed SGs with Alloy 600 tubing. (More explicit guidance on applicability constraints, voltage-based I

repair limits, and implementation actions are provided in GL 95-05.) The voltage-based repair criteri6 ensure structural and leakage integrity for all postulated design basis events. The structural criteria are intended to ensure that

~

indications subjected to the voltage repair limits will be able to withstand pressure loadings consistent with the criteria of RG 1.121. The leakage criteria ensure that for degradation subjected to the voltage repair criteria, induced leakage under the worst-case main steam line break conditions calculated using licensing basis assumptions will not result in offsite and control room dose releases that exceed the applicable limits of 10 CFR 100 and GDC 19.

Resolution: GL 95-05, " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," was issued on 08/03/95.

Comoletion Date: 08/16/95 06/30/96 A.E-28 NUREG-0933

l

[]

GCCA-0036: FAILURE OF AUTOMATIC VENTILATION SYSTEM OPERATION FOLLOWING A LOSS j

V 0F 0FFSITE POWER TAC No.: M92596

Contact:

J.R. Tappert Qgscriotion: At Waterford-3, the licensee rendered several ESF ventilation systems inoperable under certain conditions because of an engineering oversight in a design change installed between October 1992 and April 1993.

Waterford has four different ESF ventilation systems that are used to filter radioactive material from the air durino an accident. These include: (1) the shield building ventilation system, which filters the air being removed from the shield building annulus and maintains a partial vacuum inside the annulus; (2) the controlled ventilation area system, which filters air from rooms containing the emergency core cooling and containment heat removal systems in the reactor auxiliary building; (3) the control room air system (CRACS), which filters the air going into the control room; and (4) the fuel handling building ventilation system, which filters the air being removed from the fuel handling building. Each ventilation system contains two independent trains. Each train has a demister, electric heating coil, pre-filter, HEPA pre-filter, charcoal adsorber, HEPA after-filter, and 100%-capacity exhaust fan.

Automatic temperature controllers monitor the air temperature and cycle the heaters to maintain the desired air temperature. To improve the system performance and reduce system trips, the licensee developed a design change to replace the heater controllers with more accurate controllers. Because of an p;

engineering oversight, the design change caused the control circuits for the (V

ventilation system to trip the heaters after the ventilation systems lost power.

Specifically, when de-energized, the temperature controllers defaulted to the high temperature trip setpoint. Upon re-energization, the new temperature controllers responded more slowly than the 0.5 second allowed to indicate that the temperature had dropped below the high temperature condition. Following a Loss of Offsite Power (LOOP) in conjunction with a saiety injection actuation signal, the 0.5-second time-delay relay would time out and trip the ventilation heaters. With the heaters tripped, the differential temperature would decrease across the heaters and eventually trip the ventilation fan.

Oriainatina Document: Event Notification 27206, dated 05/03/94.

Reaulatory Assessment: The controller design change created a common failure mode. Three of the four ESF ventilation systems were inoperable under certain conditions from the time the design change was installed in 1993 until May 1994.

The fourth system, CRACS, was operable only because timer relays had been inadvertently left out of the preventive maintenance program and had drifted out of specification. If the fans in these ventilation systems trip, then offsite and control room radiological dose during a DBA could increase above the calculated value. Without the ventilation fans, the annulus pressure in the shield building would increase above the atmospheric pressure. Leakage from containment would be unfiltered and would release directly to the atmosphere. This direct release could have significant impact on the accident doses. However, the failure mode is limited to certain failure sequences and the heater controllers can be reset locally. The licensee calculated that operator exposure to reset the heaters Q

would be less than 1.2 rem if the LOOP was concurrent with the LOCA and less than 06/30/96 A.E-29 NUREG-0933

2 7 rem if the LOOP occurred, at a worst case, 4 days after the LOCA. An IN was proposed to notify other licensees of this event but was subsequently cancelled.

Resolution: After management review, the generic communication was cancelled based on the 8.

Grimes memorandum to NRR Division Directors,

" Advance Notification of the Division of Project Support's Intent to No Longer Support Development and Issuance of Selected Generic Letters and Information Notices,"

dated 06/10/95. The proposed IN failed to meet the threshold for regulatory action based on the following criteria- (1) length of development; (2) number of potentially affected plants; and (3) safety significance.

Comoletion Date: 07/21/95 GCCA-0037: FAILURE TO TEST SWING BUSES DURING INTEGRATED EMERGENCY DIESEL GENERATOR SURVEILLANCE TAC No.: M92597 Cont g_t.: T. Koshy Descriotion: On 09/07/94, Waterford-3 reported to the NRC that one train of safety-related equipment was inoperable since the TS-required surveillances were not performed to ensure operational readiness. The Waterford station has a third train of safety-related equipment that can be used to substitute the function of the "A" or "B" train of equipment. In the past, surveillance required by the TS was not conducted on this spare train.

Oriainatina Document: 09/07/94 Phone call to the NRC for enforcement discretion.

Reaulatory Assessment: When the NRC was made aware of this scenario, the staff reviewed and, approved the licensee's bases in continuing the operation in light of the historic performance of these components, even though a prescribed surveillance record was unavailable.

Initially, an IN was considered to promulgate the need for testing spare components that are in service to ensure its operational readiness. Later, it was recognized that the theme of this IN was contained in another recently issued IN 95-15, " Inadequate Logic Testing of Safety-Related Circuits." Therefore, an additional IN was considered unnecessary.

Resolution: None required.

Comoletion Date: 07/20/95 GCCA-0038: LIGHTNING DISSIPATION SYSTEMS TAC No.: M92599

Contact:

N.K. Hunemuller

==

Description:==

Lack of effectiveness of lightning dissipation systems in protecting plant electronics.

Oriainatina Document: None.

06/30/96 A.E-30 NUREG-0933

>%(V

)

Reaulatory Assessment: This concern was not safety-significant and was cancelled by the originator prior to any development.

Resolution: After management review, the generic communication was cancelled based on B.W. Sheron's 07/21/95 memorandum to B.K. Grimes, " Response to the Division of Project Support's Request for Review of Generic Compliance and Communications Activities (GCCA)." The proposed generic communication failed to meet the threshold for regulatory action based on its safety significance.

Completion Date: 07/21/95 GCCA-0039: SWITCHGEAR FIRE AND PARTIAL LOSS OF 0FFSITE POWER TAC No.: M92600

Contact:

E.J. Benner

==

Description:==

On 06/10/95 at Waterford-3, a generator trip occurred in response j

to failure of a lightning arrester on a remote offsite substation transformer.

The generator trip resulted in a fast transfer activation. A fire and electrical fault occurred on the 4.16kV A2 bus normal power supply breaker. The 6.9kV Al bus alternate supply breaker failed to close resulting in a loss of power to RCPs 1A and 2A. This circumstance resulted in a reactor trip and a loss of offsite power to the 4.16kV non-safety-related A2 bus and the associated 4.16kV safety-related A3 bus. An auxiliary operator informed the control room of heavy smoke within the turbine generator building. At thi.t time, the shift supervisor did not activate rm the plant fire alarm or dispatch the fire brigade, but directed two auxiliary

!]

operators to don protective gear and investigate whether a fire existed. The

\\

operators reported seeing flames above the A2 switchgear and the shift supervisor activated the fire brigade. Operators requested assistance from the local offsite fire department. The fire brigade was unable to suppress the fire using portable fire extinguishers. The offsite fire department arrived on the scene and extinguished the fire with water after the A2 bus was deenergized. A reflash occurred which had to be put out with water a second time. During the cooldown transition from Mode 4 to Mode 5, operators discovered that the isolation valves for both trains of shutdown cooling did not operate properly. The plant cooldown to Mode 5 was delayed approximately 38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br /> while these valves were repaired.

Oriainatina Document: Event Notification 28923, dated 06/10/95.

Reaulatory Assessment: The NRC conducted an augmented inspection team (AIT) inspection to determine the causes, conditions, and circumstances relevant to this event. The results of this AIT inspection are documented in NRC Inspection Report 50-382/95-15, dated 07/07/95. The AIT identified three primary concerns:

fire protection, fast bus transfer design, and shutdown cooling valve inoperability. Evaluation of these three concerns by NRR revealed that no immediate generic safety issue existed and that dissemination of the information to the industry would be an appropriate technical resolution.

Resolution: IN 95-33, "Switchgear Fire and Partial loss of Offsite Power at Waterford Generating Station, Unit 3," was issued on 08/23/95.

O Completion Date: 08/23/95 V

06/30/96 A.E-31 NUREG-0933

GCCA-0040: SPENT FUEL TRANSFER CANAL SHIELDING DEFICIENCY AT BOILING WATER REACTOR TAC No.: M92876

Contact:

T.A. Greene

==

Description:==

In November 1994, contractors at Hatch were conducting underwater operations in the Unit I spent fuel pool - cutting coupons out of spent control rod blades containing the upper guide roller bearings. The highly activated stellite bearings (some measured as high as 160 sievert [16,000 rem] per hour at 30 centimeters (12 inches] under water) were being collected adjacent to the Unit I work area. Periodically, the coupons containing the upper guide roller bearings were transferred from the collection bucket to a cask liner in the shipping cask storage area in the Unit 2 spent fuel. When the workers could not find the tool to open the liner, they decided to transfer the coupons in the collection bucket (about half full, with 160 coupons) into another bucket for temporary storage, so that the cutting process could continue. To facilitate the task, the off-load was performed in the transfer canal since the canal was much shallower than the fuel pools. During the transfer, some of the coupons fell to the bottom of the transfer canal. Since the transfer canal was designed and routinely used as a transit area for highly activated material, including spent fuel, the workers were not concerned about dropping the coupons and saw no need to notify the unit shift foreman or the control room of the incident. They recovered the coupons from the bottom of the canal and placed them in the storage bucket resting on the canal floor.

About 30 minutes after the coupons were dropped, a plant operator was walking through the Unit I hallway directly under the transfer canal when his digital alarm dosimeter alarmed on high dose rate. The plant operator left the area promptly and notified a health physics supervisor. The licensee measured radiation levels of up to 1 sievert [100 rem] per hour on contact with the hallway ceiling directly below the bottom of the canal and 0.05 to 0.1 sievert

[5 to 10 rem] per hour in the general area of the hallway.

Oriainatina Documents:

Inspection Reports 50-321/95-01 and 50-366/95-01

[ Accession No. 9502140081].

Reaulatory Assessment: The plant equipment operator received a dose equivalent of about 0.1 msievert [10 mrem] from the event, which is below the regulatory limit. The licensee has always required that all personnel entering the radiologically-controlled area be issued dosimeters. Before alarm dosimeters were required, all workers entering the radiologically-controlled area were issued personnel dosimeters (non-alarm), so any doses to workers from this shielding deficiency would have been detected and accounted for as part of the routine dosimetry program.

Many licensees perform sent fuel pool modifications and major cleanup activities involving handling and moving large quantities of highly activated materials, including spent fuel. In general, the industry has significantly improved its awareness of and controls for potentially high and very high radiation areas caused by operational mishaps (e.g., dropping a spent fuel bundle in the transfer canal directly over the up*per drywell in a BWR, and the hazards of withdrawn in-core thimbles under the reactor vessel at PWRs). Initiatives taken by licensees should help prevent unexpected, uncontrolled worker exposures with the potential for exceeding the regulatory limits. A thorough pre-job evaluation of activities 06/30/96 A.E-32 NUREG-0933

O>

involving highly activated (or potentially highly activated) components can help y

identify challenges to existing plant shielding.

Resolution: At Hatch, the licensee instituted procedural control to prohibit the use of the transfer canal until doors had been installed in order to exclude worker access to the affected hallway. Before the transfer gates are allowed to i

be lifted (allowing use of the transfer canal), the doors at each end of the hallway under the canal are locked and access to the hallway is controlled as a very high radiation area. IN 95-56, " Shielding Deficiency in Spent Fuel Transfer Canal at a Boiling-Water Reactor," was issued on 12/12/95 to alert the industry.

i Comoletion Date: 12/11/95 GCCA-0041: CilANGES IN THE OPERATOR LICENSING PROGRAM TAC No.: M92877

Contact:

J.W. Shapaker Descrint_ign: The NRC intends to change the operator licensing process to give facility licensees the option of preparing written examinations and operating tests for operator and senior operator license applicants. The NRC will review and approve licensee-proposed examinations and tests and will independently conduct the operating tests. Facility licensees will only conduct the written examinations. The NRC will review the graded written examinations, grade each applicant's operating test performance, make the final pass or fail decisions, and issue licenses, as appropriate. From October 1995 to March 1996, a voluntary g\\

pilot progren to evaluate and refine the new examination development process will

/V be conducted.

Oriainatina Documents: NUREG-1021, Revision 7,

" Operator Licensing Examiner Standards"; NUREG/BR-0122, " Examiners' Handbook for Developing Operator Licensing Written Examinations"; GL 95-06, " Changes in the Operator Licensing Program."

Reaulatory Assessment: 10 CFR 55, " Operators' Licenses," establishes NRC procedures and criteria for issuing licenses to operators and senior operators.

It does not, however, define a specific process for conducting licensing examinations. Guidance in this area is given in NUREG-1021, " Operator Licensing Examiner Standards," which includes the procedures that NRC staff examiners and NRC-certified contract examiners use to prepare and conduct both the written and operating portions of the licensing examinations. The role of power reactor facility licensees had been limited to reviewing and validating the NRC-prepared examinatic,ns before they are given, and to providing administrative and i

logistical support to the NRC and contract examiners while the examinations are in progress. Over the past 10 years, power reactor licensees have placed increased emphasis on establishing accredited training programs and, as a result, improvements in operator training and performance have been evident in the initial operator licensing process. This fact, in conjunction with a continuing effort on the part of the NRC to streamline the functions of the Federal Government consistent with White House initiatives and to accommodate anticipated 1

resource reductions, motivated the NRC to reconsider its approach to conducting the initial operator licensing examination program. The NRC intends to change the q

guidance in NUREG-1021 to permit facility licensees greater participation in the j

initial operator licensing program.

06/30/96 A.E-33 NUREG-0933

Resolution: GL 95-06, " Changes in the Operator Licensing Program," was issued on 08/15/95.

Comoletion Date: 08/16/95 GCCA-0042: UNPLANNED. UNMONITORED RELEASE OF RADI0 ACTIVITY FROM THE EXHAUSl YENHLATION SYSTEM OF &_M TAC No.: M92935

Contact:

C.V. Hodge Descriotion: On 04/05/95, an unplanned, unmonitored release of high specific activity-mixed radioactive corrosion products was released from the south plant vent of Hope Creek. The licensee believed that no release had occurred because of a lack of unusual indications. Large portions of the protected area were contaminated, including onsite vehicles and buildings, and contamination levels accessible to personnel was about 10' dpm/100cm'. One vehicle, located downwind of the release point, left the site and was determined two days later to be contaminated. The radioactive contamination came from a small capacity evaporator for decontamination solution discharging directly to the south plant vent in the turbine building. The licensee believed, from lack of unusual indications, that no release had occurred. About 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> into the event, the licensee observed high contact dose rates from effluent ducts and highly contaminated liquid leaking from the ducts. The leaking liquid was attributed to a pre-existing condition.

Oriainatina Document: Inspection Report 50-354/95-05.

Reaulatory Assessment: Region I conducted a special inspection and found that:

(1) no release limits were exceeded; (2) worst-case analyses indicated that the release had little radiological effect on the public and environment; (3) safe operation of the reactor was not affected; (4) the release did not significantly affect onsite personnel; and (5) the licensee did an excellent evaluation of the environmental effect of the release. However, Region I concluded that: (1)there were four apparent violations for escalated enforcement existed related to (a) lack of adequate approved operating procedures for this evaporator (TS 6.8.1),

(b) inadequate monitoring to enable detection of the release (10 CFR 20.1501(a)),

(c) alarm setpoint changes not made in accordance with approved procedures (TS 6.8.1),

and (d) workers not being informed of the release and onsite contamination once it was identified (10 CFR 19.12); (2) there was a particular concern about the inability of licensee personnel to readily determine that a release had occurred; and (3) there was a particular concern that licensee design reviews did not identify the potential for an unmonitored release, despite previous NRC-published information on this subject such as (a) IN 91-40,

" Contamination of Nonradioactive System and Resulting Possibility for Unmonitored, Uncontrolled Release to the Environment," dated 06/19/91, (b)

Circular 80-18, "10 CFR 50.59 Safety Evaluations for Changes to Radioactive Waste Treatment Systems," dated 08/22/80, and (c)Bulletin 80-10, " Contamination of Nonradioactive System and Resulting Potential for Unmonitored, Uncontrolled Release of Radioactivity 05/06/80. This particular event represents minimal,to the Environment," dated safety significance; however, its potential safety significance and potential generic implication support the Regicn I

recommendation, endorsed by TERB/D0TS staff, to issue the proposed IN.

06/30/96 A.E-34 NUREG-0933

Resolnijsn: IN 95-46, " Unplanned, Undetected Release of Radioactivity from the Exhaust Ventilation System of a Boiling Water Reactor," was issued on 10/06/95.

Completion Date: 10/06/95 GCCA-0043: SUSCEPTIBILITY OF LOW PRESSURE COOLANT AND CORE SPPAY INJECTION VALVES TO PRESSURE LOCKI$

Contact:

T.A. Greene Descriotion: From May to July 1995, Georgia Power Company, the licensee for Hatch-1 & 2, experienced several valve failures in both units during testing of its in-board LPCI 24-inch flexible gate valve. During the investigation of the root causes for these failures and in responding to the NRC staff's inquiries, the licensee realized that, because of back-leakage of the downstream check valves, reactur system pressure could cause the LPCI valves and the core spray injection valves to be susceptible to pressure locking during an accident.

Oriainatina Document: Inspection Report 50-321/95-17, dated 08/24/95.

Reaulatory Assessment: The NRC staff and the nuclear industry have been aware of pressure-locking problems in gate valves for many years. The industry has issued several event reports describing failures of safety-related gate valves to operate because of pressure locking. Several generic industry communications have O

given guidance for identifying susceptible valves and for taking appropriate preventive and corrective measures. Also, the NRC staff has provided information on pressure locking of gate valves to the industry and has discussed the safety significance of the potential for pressure locking of gate valves at public meetings.

Resolution: The generic concern was resolved by the issuance of IN 95-30,

" Susceptibility of low-Pressure Coolant Injection and Core Spray Injection Valves to Pressure Locking," and GL 95-07, " Pressure Locking and Thermal Binding of Safety-Related Power-0perated Gate Valves."

Comoletion Date: 08/03/95 GCCA-0044: POTENTIALLY NONCONFORMING FASTENERS SUPPLIED BY A&G ENGINEERING II.

LL TAC No.: M93226

Contact:

J.R. Tappert

==

Description:==

IN 95-12, "Potentially Nonconforming Fasteners Supplied by A&G Engineering II, Inc.," was issued on 02/21/95 to alert licensees to potentially non-conforming fasteners supplied from A&G Enginu "ng II, Inc. Additional information concerning de-carburization and the manufacturer of the fasteners was reviewed after IN 95-12 was issued. A supplement was proposed to relay this additional information to licensees.

O Oriainatina Document: 01 Report 5-93-016R re A&G Engineering II, Inc.

06/30/96 A.E-35 NUREG-0933

Reaulatpry Assessment: PEC0 found that 143 out of 150 heavy hex nuts supplied by A&G and an intermediary vendor HUB, Inc. had carbon contents significantly different from that reported on the Certified Material Test Report (CMTR). The nuts were certified by A&G to be traceable to Hamanaka Nut Mgn. Co. Ltd. The NRC has reviewed additional information since the issuance of IN 95-12. Further analysis of the rejected nuts from PEC0 found that, when the surface material was removed, the remaining metal had the proper carbon content. This would indicate that surface decarburization was primarily responsible for the previously identified low carbon values. However, it is not clear what effect the observed surface decarburization has on the thread integrity of the nuts. The NRC has also been contacted by the Hamanaka Nut Manufacturing Co., Ltd. Hamanaka believes that the rejected nuts were fraudulently labeled with their trademark. Hamanaka hot-forms their trademark (raised), whereas the 143 rejected nuts had a stamped Hamanaka marking.

This concern is generic because A&G had supplied a significant amount of safety-related threaded fasteners to the nuclear power industry. The proposed IN supplement updates the industry on the recent information concerning surface decarburization and the Hamanaka trademark issues. The actual safety significance of the surface decarburization has not been ascertained. The larger issue of substandard / nonconforming materials being supplied to utilities is being addressed by the Special Inspection Branch (TSIB). They are performing a number of inspections of vendors and utilities to determine the scope of the problem.

Additionally, RES has been requested to determine what level of sampling is required for commercial grade dedication. TSIB proposes to issue a generic communication in the near future documenting their findings.

Resolution: Supplement I to IN 95-12 was issued on 10/05/95 to alert licensees to relevant plant operating experiences.

Completion Date: 10/05/95 GCCA-0045: CURRENT FIRE ENDURANLi TEST RESULTS FOR 3M INTERAM RACEWAY FIRE BARRIER SYSTEMS TAC No.: M93295

Contact:

T.J. Carter Descriotion: A number of full-scale fire endurance tests were sponsored by Peak Seals Corporation. The results of the fire endurance tests for electrical raceway fire barrier systems constructed from 3M Company Interam fire barrier materials have not been provided to licensees.

Oriainatina Document: None.

Reaulatory Assessment: The staff was informed about the test program and witnessed the tests. To provide information to licensees as soon as possible, the staff proposed that an IN be issued to present their observations. The staff has not reviewed the test reports.

Resolution: IN 95-52, " Fire Endurance Test Results for Electrical Raceway Fire Barrier Systems Constructed from 3M Company Interam Fire Barrier Materials," was issued on 11/14/95 to inform licensees of fire endurance test results.

06/30/96 A.E-36 NUREG-0933

I J

Comoletion Date: 11/14/95 GCCA-0046: P0TENTIAL FOR DATA COLLECTION EQUIPMENT TO AFFECT PROTECTION SYSTEM PERFORMANCE TAC No.: M93359

Contact:

E.M. McKenna Descriotion: Quad Cities-2 found that a portable computerized data acquisition j

system (DAS), connected to instrumentation circuits to collect data, was affecting the signals being monitored. I'wo different problems occurred:

i I

(1)

In December 1994, the licensee found that a RPS setpoint on low water level was reading out of tolerance with the DAS connected to the circuit, i

but de-energized. Further evaluation found that in this configuration, the 1

DAS was a low internal impedance path in the circuit, allowing more current to be drawn and decreasing the signal. This problem is similar to 1

l an event at Fermi-2 (that occurred in February 1995) that was the subject j

of IN 95-13.

1 (2)

In July 1995, the licensee re-installed the DAS to collect further data.

Instructions and precautions were taken for the system not to be de-energized while connected (to avoid the above problem). However, two i

unexpected interactions of the energized DAS with the circuits being 1

monitored occurred. First, there was a change in recirculation pump speed

o (in manual control) with no change demanded by the operator. Later, one

) (\\

water level channel was indicating 8 inches low (and there were also 1

changes in total indicated core flow), The DAS was connected to the i

circuits being monitored by a ribbon cable that was unshielded from a terminal block external to the DAS to a circuit board within the DAS. The l

signals being processed are DC signals. Pre-installation testing had shown i

that stray AC voltage signals (present because of AC power supply to i

circuits) were present in the circuit. This AC voltage normally did not i

affect proper functioning of the circuits. When the DAS was in use, investigation revealed that electromagnetic interference across circuits l,

within the ribbon cable caused feedback from one circuit into another.

This feedback would result in a signal that would add or subtract from the i

desired signal being processed and resulted in changes in both control (recirculation pump) and indication.

Oriainatina Document: Morning Report 3-95-0130, dated 08/07/95.

j Reaulatory Assessment: The DAS unit was considered by the licensee to be non-J intrusive, so detailed testing of the possible effects of the unit on parameters being monitored was not performed. Redundant channels of instrumentation were i

simultaneously connected to the DAS, compromising the electrical separation of independent channels that is required. Most, but not all, of the channels affected at Quad Cities were non-safety related circuits. Only small changes in signals resulted, which were detected by plant personnel, and redundant channels s

were not simultaneously affected. An IN had been issued on the Fermi event (where i

the DAS was de-energized). la reinforce the need to consider possible interactions of a DAS, (especially since it may be connected to redundant 4

channels), a supplement to the IN was issued.

J 06/30/96 A.E-37 NUREG-0933

l l

Resolution: IN 95-13, Supplement 1, " Potential for Data Collection Equipment to Affect Protection System Performance," was issued on 11/22/95 to alert licensees l

to the relevant plant operating experience, but does not require any licensee action.

Comoletion Date: 11/22/95 I

GCCA-0047: BORAFLEX DEGRADATION IN SPENT FUEL POOL STORAGE RACigi TAC No.: M93373

Contact:

P.C. Wen Descriotion: At South Texas Project (STP), blackness (neutron absorption) testing was perfcrmed during August 1994 on selected Unit 1 spent fuel pool storage racks I

to determine the condition of the Boraflex. The test results indicated that the Boraflex had significantly degraded due to dissolution of the Boraflex in the spent fuel pool environment.

Oriainatina Documenti: Event Notification 28277 and LER 50-498/95-002.

Reaulatory Assessment: On the basis of test and surveillance information from plants that have detected areas of Boraflex degradation, no safety concern exists that warrants immediate action. Boraflex dissolution appears to be a gradual and localized effect forewarned by relatively high' silica levels in the pool water.

Because of the safety margin present in spent fuel storage pools, compliance with the required sub-criticality margin (or conformance with the same margin to which licensees have committed in their updated FSARs) can be expected to be maintained during the initial stage of Boraflex degradation. This safety margin is due to the 5% sub-criticality margin assumed in the analysis, the generally lower reactivity of stored fuel than that assumed in the safety analysis, and, in the case of PWRs, the presence of borated water in the pool. Therefore, continued facility operation is justified.

Boraflex degradation is a generic concern. Since there are over 50 pools that contain Boraflex and in light of the problems that were discovered at Palisades (IN 93-70), STP, and most recently at Fort Calhoun, this Boraflex concern may be a widespread problem.

The staff proposed that an IN be issued to inform licensees of this concern and that a GL be developed to require appropriate actions. TAC No. M93373 was initiated for the development of the IN.

Resolution: IN 95-38, " Degradation of Boraflex Neutron Absorber in Spent Fuel Storage Racks," was issued on 09/08/95. A GL on this topic is being developed by SRXB.

Comoletion Date: 09/08/95 l

I 06/30/96 A.E-38 NUREG-0933 1

i I

i GCCA-0048: RISK IMPACT OF MAINTENARCE DURING LOW POWER OPERATION AND SHUTDOWN TAC No.: M93642

Contact:

N.K. Hunemuller Descriotion: The details of a study were published which evaluated the risk impact of LCOs at low power and shutdown in the current TS for Grand Gulf. A probabilistic model was developed for each of eight plant operational states. The study indicated that the increase in conditional CDF for taking a single train 4

of standby service water out of service during low power and the first few days of hot shutdown was comparable to that at ful

?ower operation. During hot shutdown and cold shutdown, the study indicated.nat the increase in conditional CDF could exceed that at full-power operation.

Oriainatino Document: NUREG/CR-6166, " Risk Impact of BWR Technical Specifications j

Requirements During Shutdown," published October 1994.

Reaulatory Assessment: On 05/23/95, the Generic Issues and Events Assessment Panel initially determined that there was a lack of sufficient data to support a conclusion of generic applicability on the specific issue of the impact of service water maintenance shortly after shutdown. However, the broad concern of 1

low power and shutdown risk was generic. Although NUREG/CR-6166 was a public.

l document, there was sufficient management interest to proceed with development of an IN to enture that licensees were aware of the study. After further review, the Panel authorized development of an IN on 09/12/95.

Resolution: IN 95-57, " Risk Impact Study Regarding Maintenance During Low-Power O

Operation and Shutdown," was issued on 12/13/95 to inform licensees of an RES study on the effect of system or equipment maintenance on BWR low-power and shutdown CDF, but no licensee action was required.

Comoletion Date: 12/13/95 StCCA-0049: LEGAL ACTIONS AGAINST THERMAL SCIENCE. INC.. MANUFACTURER OF THERMO-LAG TAC No.: M93668

Contact:

T.J. Carter

==

Description:==

An IN was previously issued to inform interested parties that there was an indictment pending against Thermal Science, Inc. A jury subsequently acquitted Thermal Science of all charges.

QClainatina Document: IN 94-86, " Legal Actions Against Thermal Science, Inc.,

Manufacturer of Thermo-Lag," issued 12/22/94.

Reaulatory Assessment: This is not a safety issue. However, there is a need to clarify the record presented by the original IN.

Re1919 tion: IN 94-86, Supplement 1, " Legal Actions Against Thermal Science, Inc.,

Manufacturer of Thermo-Lag," was issued on 11/15/95 to inform licensees that a jury acquitted Thermal Science, Inc. and its president of making false

(~N statements.

[ompletion Date: 11/15/95 06/30/96 A.E-39 NUREG-0933

GCCA-0050: TRANSIENT INVOLVING OPEN SAFETY RELIEF VALVE FOLLOWED BY COMPLICATIONS TAC No.: M93705

Contact:

T.J. Carter

==

Description:==

During routine operation of Limerick Unit 1 on 09/11/95, one 2-stage Target Rock SRV opened and stuck open causing a blowdown transient. This was the first time this particular type of SRV had unexpectedly opened; cause was attributed to excessive steam leakage through the pilot valve. Reason for the initial leakage is not known, but it is known that the stellite seat and disk had ultimately become badly eroded. Suppression pool cooling was ongoing to remove heat being added to the pool by the leaking steam. However, as a result of the blowdown transient, additional pool cooling was required and a second pump was started. Erratic flow was observed subsequently on the initially running pump.

This was interpreted as pump cavitation caused by debris clogging the suction strainer. Upon inspection, the suction strainer on the pump that exhibited erratic performance was found to be largely covered with a fibrous material and sludge.

Oriainatina Document: Event Notification 29316, dated 09/11/95.

Reaulatory Assessment: Sudden opening of an SRV is an initiating event that challenges safety systems and has been analyzed. However, excessive steam leakage past the pilot valve with the subsequent SRV opening is new and raises a generic concern since licensees cannot differentiate steam leaking through the main valve or the pilot valve. The previously observed accumulation of material on the suction strainers at several reactors has already led to a generic action plan which is still in progress. This event itself is of minor immediate concern since the cooling function was retained. However, the fact that the licensee had been operating for months with leaking SRVs is indicative of a potential " work around" with unexpected consequences. Furthermore, the suppression pool at Limerick Unit I had never been cleaned even though an opportunity existed after the licensee had removed a substantial amount of material from the Unit 2 pool during its cleaning. This may demonstrate another lack of appreciation of the potential consequences. An IN was prepared rapidly to alert licensees to the potential precursor of spurious SRV opening and reiterate the concern about unclean suppression pools. The degradation of strainer performance due to debris accumulation during normal operation may require additional generic action.

Resolution:

IN 95-47,

" Unexpected Opening of a Safety / Relief Valve and Complications Involving Suppression Pool Cooling Strainer Blockage," was issued on 10/04/95 to alert licensees to a failure of an SRV to remain closed during steady state operation and the ensuing complications involving blockage of a strainer located in the suppression pool.

Completion Date: 10/04/95 0

06/30/96 A.E-40 NUREG-0933

O GCCA-0051: UNEXPECTED CLOGGING OF RHR PUMP STRAINER WHILE OPERATION IN SUPPRESSION POOL COOLING MODE TAC No.: M93740

Contact:

J.W. Shapaker

==

Description:==

On 09/11/95, Limerick-1 experienced a stuck-open SRV while operating at 100% power. Attempts to close the valve were unsuccessful and a manual reactor scram was initiated. The licensee initiated suppression pool cooling and subsequently experienced clogging of one RHR pump suction strainer.

Oriainatina Document: Event Notification 29316, dated 09/11/95.

Reaulatory Assessment: 10 CFR 50.46 requires that licensees design their ECCS so that the calculated cooling performance following a LOCA meets five criteria, one of which is to provide long-term cooling capability of sufficient duration following a successful system initiation so that the core temperature shall be maintained at an acceptably low value, and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core. The ECCS is designed to meet this criterion, assuming the worst single active failure and only partially obstructed flow through the strainer (s). The Limerick event demonstrated that inadequate suppression pool cleanliness can adversely impact ECCS performance and could prevent the ECCS from performing its safety function of long-term decay heat removal following a LOCA. As a result, BWR licensees were requested to take certain actions that, if required, would be compliance backfits under the terms of 10 CFR 50.109(a)(4)(1),

i.e., would be p

necessary to ensure compliance with NRC rules and regulations. IN 95-47, D)

" Unexpected Opening of a Safety / Relief Valve and Complications involving Suppression Pool Cooling Strainer Blockage," was issued on 10/04/95 to alert licensees about the event and its consequences.

Resolution: Bulletin 95-02, " Unexpected Clogging of a Residual Heat Removal (RHR)

Pump Strainer While Operating in Suppression Pool Cooling Mode."

Conoletion Date: 11/26/95 GCCA-0052: UNEXPECTED OPENING OF AN SRV AND COMPLICATIONS INVOLVING SUPPRESSION POOL STRAINER BLOCKAGE TAC No.: M93840 Sontact: E.M. McKenna Descriotion: On 09/11/95, an SRV at Limerick-1 opened unexpectedly. An IN issued on 10/04/95 to alert licensees about the event required revision.

j Oriainatina Document: IN 95-47, " Unexpected Opening of a Safety / Relief Valve and Complications Involving Suppression Pool Cooling Strainer Blockage," dated 10/04/95.

Reaulatory Assessment: This TAC was opened for issuance of a revision to IN 95-

47. (See GCCA-0050 for further details on the initiating event and the IN.) In addition,Bulletin 95-02, " Unexpected Clogging of RHR Pump Strainer While

[,]

Operating in Suppression Pool Cooling Mode," was issued on 10/17/95 (See TAC V

M93740).

06/30/96 A.E-41 NUREG-0933

i The IN revision was proposed to clarify licensee planned actions when tail pipe temperature limits were reached and to note that the NRC plans to evaluate the potential generic implications of significant leakage through Target Rock 2-stage safety / relief valves, such as occurred at Limerick (this generic effort is being tracked under TAC No. M93841). Details on licensee proposed action plans are contained in a letter dated 10/06/95, from PECO Energy Company [ Accession No.

9510120218].

Resolution: IN 95-47, Revision 1, " Unexpected Opening of a Safety / Relief Valve and Complications Involving Suppression Pool Cooling Strainer Blockage," was issued on 11/30/95.

Completion Date: 11/30/95 l

GCCA-0053: DECAY HEAT MANAGEMENT PRACTICES DURING REFUELING TAC No.: M94087

Contact:

D.L. Skeen Descriptio2: Adequacy of decay heat management practices during refueling outages was called into question as a result of an NRC review of a design change at Millstone-1 (submitted to the staff on 07/28/95). The practice of unloading the entire core from the reactor vessel to the SFP during refueling has become fairly common. However, a full core off-load during refueling may potentially exceed the ability of the SFP cooling system to maintain SFP temperature within its design limit, if other means are not available to assist the SFP cooling system.

The staff's guidance for review of SFP cooling system design in SRP Section 9.1.3 I

specifies consideration of a single failure of a cooling system component in evaluating the capability for long-term cooling of the SFP. However, the guidance specifies that the evaluation of cooling system capability for short-term, high heat load conditions need not consider a single failure of a cooling system component. Regardless of this guidance, the licensing basis for SFP cooling is not consistent from plant to plant.

Oriainatina Document: LER 50-245 93-011, dated 10/18/93.

Reaulatory Assessment: Not all plants perform a full-core off-load during refueling. Some plants that do perform full-core off-loads have adequately addressed them in their FSARs. Plants are most at risk shortly after shutdown when decay heat is at its highest level. After some time has past (in the range of a few days) the decay heat has had a chance to dissipate to the point where SFP cooling systems can handle it. Most plants do not start moving fuel for at least a few days after shutdown. Also, although the design temperature limit for a SFP may be 150 F, administrative limits in the range of 110 to 125 F are in place to protect personnel working on the refueling floor. If temperatures were to get into this administrative range during refueling, the core off-loading could simply be stopped until additional cooling could be aligned or the heat dissipated over time.

Resolution: On 11/21/95, NRC determined that development of an IN was warranted.

The notice alerted licensees to the importance of assuring that: (1) planned core off-load evolutions, including refueling practices and irradiated decay heat removal, are consistent with the licensing basis, including the FSAR, TS, and 06/30/96 A.E-42 NUREG-0933

~. _ _... _. _ _ _

license conditions; (2) changes are evaluated through the application of the provisions of 10 CFR 50.59, as appropriate; and (3) all relevant procedures associated with core offloads have been appropriately reviewed.

IN 95-54, " Decay Heat Management Practices During Refueling Outages," was issued on 12/01/95. The staff continues to look at the adequacy of SFP cooling systems and other cooling systems that are available during refueling as part of an action plan'(TAC No, M88094) that began when questions were raised about the design basis of the SFP at Susquehanna Steam Electric Station.

Comoletion Date: 12/01/95 GCCA-0054: P0TENTIAL FOR LOSS OF AUTOMATIC ENGINEERED SAFETY FEATURES ACTUATION TAC No.: M94179

Contact:

C. Doutt

==

Description:==

On 02/02/95, during performance of an analysis of the potential effects of a steamline break such as pipe whip and jet impingement in the turbine building on the automatic ESF functions of the W SSPS, Pacific Gas and Electric Company (PG&E), the licensee for Diablo Canyon, identified a condition where such a break could result in failure of one train of the SSPS. This failure concurrent with a single active failure in the other SSPS train (the plant licensing basis) could result in loss of the automatic SSPS ESF functions needed to mitigate the steamline break.

b Oriainatina Documents: 10 CFR 50.72, Event Notification 28318, dated 02/02/95, O

and IN 95-10, Supplement 1, " Potential for Loss of Automatic Engineered Safety Features Actuation," dated 02/10/95.

Reaulatory Assessment: Subsequent to notification of the concern by PG&E, the licensees for Farley, North Anna, Salem, Sequoyah, Shearon Harris, and Summer informed the staff that their SSPS was subject to the same concerns as those noted at Diablo Canyon. Based on the low probability of the steanline break of concern, and the fact that the manual ESF actuation capability was not affected by the postulated steamline break, the staff granted enforcement discretion for a period of two weeks to those licensees who declared the SSPS inoperable, in order for them to make the necessary modifications to fix the problem.

Resolution: In order to meet the plant licensing basis for compliance with the single failure criterion as stated in IEEE 279, the affected licensees modified the SSPS circuitry by providing qualified electrical isolation of the SSPS power supply from the input circuitry. The staff confirmed that affected licensees completed and satisfactorily tested the modification. In addition, the staff reviewed the licensing basis for other W plants with the SSPS and confirmed that they are not subject to the postulated failure mode identified at Diablo Canyon.

Comoletion Date: 04/11/95 0

06/30/96 A.E-43 NUREG-0933

.~.

-_ -, U

GCCA-0055: FAILURES IN ROSEMOUNT PRESSURE TRANSMITTERS DUE TO HYDROGJJ EEBMEATION It!TO THE SENSOR CELL TAC No.: M91644

Contact:

S. V. Ath tvale Descriotion: On 11/22/94, while St. Lucie was in cold shutdown and in the process of filling and venting the RCS, two of the four pressurizer pressure channel transmitters failed high within 5 minutes of each other resulting in initiation of safety injection on high pressure signals. The affected transmitters were manufactured by Rosemount, Inc. Subsequent analysis of the failed transmitters identified entrapped hydrogen gas in the sensor cell which caused the sensor cell isolator diaphragm to deform (bow). Additional analysis determined that the isolating diaphragm material was made of Monel Alloy 400 instead of the Rosemount specified 316L stainless steel for this service. Monel 400 is permeable to monatomic hydrogen.

Oriainatina Documents: LER 50-335,94-009, dated I?/19/94; Rosemount 10 CFR 21, dated 03/21/95; IN 95-20, " Failures in Rosemount Pressure Transmitters Due to Hydrogen Permeation into the Sensor Cell," dated 03/22/95.

Reaulatory Assessment: The staff determined that the identified failure in Rosemount transmitters was generic, but the Rosemount Part 21 indicated only a relatively small number of affected transmitters made from a lot of Monel 400 rather than 316L stainless steel. Further, the affected transmitters are for use in high pressure systems only. No other licensees identified similar failures in Rosemount transmitters and, because the transmitters fail high, the failure is readily detectable during calibration and would be indicated to the operator in the control room. in addition, the failure was likely the result of gas evolution from depressurization and re-pressurization during shutdown and startup and, therefore, would not occur during normal power operation.

The staff requested all licensees to address this concern by: (1) determining whether affected Rosemount transmitters were being used in their plants; (2) assessing the impact of these transmitters on plant operational safety; and (3) identifying when the affected transmitters would be replaced with the properly designed transmitters. The staff reviewed the licensee responses and determined that proper actions were being taken to deal with the concern.

Easolution: Licensees committed to replace the affected Rosemount transmitters with correctly designed transmitters or confirmed that no affected transmitters were used in a service where they could impact plant safety. The staff conducted an inspection of the Rosemount QA process in order to confirm that a subsequent QA breakdown, such as the material error, will not occur again. The staff verified that Rosemount has taken the necessary steps to correct the QA process.

Completion Date: 04/25/95 FCA-0056: AUGMENTED REACTOR VESSEL INSPECTION TAC No.: M67462

Contact:

E.J. Benner

==

Description:==

10 CFR 50.55a (g)(6)(ii)(A), " Augmented Examination of Reactor Vessel," was issued in 1992 to require a one-time augmented inspection of the 06/30/96 A.E-44 NUREG-0933

i 1

1 IO reactor vessel in accordance with the 1989 Edition of Section XI of the ASME i V Code. TAC No. M67462 was issued to DE to answer questions that licensees had concerning the new rule.

Oriainatina Document: Issuance of new Rule 10 CFR 50.55a(g)(6)(ii)(A), Federal i

Register Notice 57 FR 34673, dated 08/06/92.

Reaulatory Assessment: Based on interactions between DE and the industry on j

related ISI reviews, it became apparent that several licensees were either l

urware of the rule or had misconceptions regarding staff approvals required by 1

the rule when complete inspection coverage cannot be achieved. DE subsequently requested an additional TAC No. (M93643) to develop an IN to alert licensees to i

the presence of the rule and to provide clarification of the rule to licensees.

i' An IN is an appropriate generic resolution. Greater generic action is not warranted because of the pre-existing presence of the rule. The IN is warranted j

because of evidence that licensees are either unaware of the rule or unaware of the full ramifications of the rule demonstrated by a lack of comprehensive examinations by licensees.

Resolution: IN 96-32, " Implementation of 10 CFR 50.55a(g)(6)(ii)(A), ' Augmented Examination of Reactor Vessel,'" was issued on 06/05/96 (Accession No.

9605200277).

Comoletion Date: 05/29/96 GCCA-0057: CONSIDERATION OF POSITION CHANGEABLE VALVES TAC No.: M82072

Contact:

J.W. Shapaker Descriotion: The NRC, with the assistance of BNL, reviewed and evaluated the concerns associated with the mis-positioning of valves from the control room and determined that the recommendations in GL 89-10, " Safety-Related Motor-0perated Valve Testing and Surveillance," should be changed.

Oriainatina Document: Letter from the H Owners Group, dated 07/21/92, requesting that the NRC notify PWR licensees that the provisions of GL 89-10 for valve mis-positioning are not applicable to PWRs.

Reaulatory Assessment: The NRC no longer considers the inadvertent operation of MOVs from the control room to be within the scope of GL 89-10 for PWRs.

Therefore, Supplement 7 to GL 89-10 is a relaxation of the recommendations set forth in GL 89-10 and prior supplements. Implementation of this relaxation is voluntary and Supplement 7 requests neither actions nor information from licensees. Licensees that may have taken action, or made commitments related to valve mis-positioning prior to the issuance of Supplement 7, are allowed to take advantage of this relaxed position provided licensees document the change in i

their GL 89-10 program.

Resolution: Issuance of GL 89-10, Supplement 7,

" Consideration of Valve O

Mispositioning in Pressurized-Water Reactors," dated 01/24/96 (Accession No.

9601190442).

06/30/96 A.E-45 NUREG-0933

Completion Date: 01/26/96 GCCA-0058: PROBLEM OF GREASE LEAKAGE IN PRE-STRESSED CONCRETE CONTAINMENT TAC No.: M85236

Contact:

T.A. Greene

==

Description:==

There are 41 pre-stressed concrete containments (PCC) with greased unbonded tendons in the U.S. The ISI requirements for PCCs provide an assurance that the grease leakage will not result in inadequate protection of tendon elements against corrosion. However, there is a concern that, if the petroleum-based grease leaks into the concrete constituents in significant amount, the concrete strength properties (compressive, shear, and bond strengths) could be reduced to an extent that the containment's capacity is appreciably dcgraded.

The purpose of this TAC was to investigate the properties of concrete in PCC as affected by the permeation of grease.

Oriainatina Document: NRR became aware of significant grease leakage through the concrete of several PCCs, e.g., Trojan, ANO-1, TMI-1, and Fort Calhoun.

Reaulatory Assessment: NRR is increasingly examining the effectiveness of its regulations from a risk perspective. Current regulations that govern containment design and performance are derived from assuring that the containment will withstand DBEs and provide for structural margin in the design. When evaluated under severe accident conditions, both reinforced and pre-stressed concrete containments are calculated to have ultimate failure pressure two to three times the design pressure. The results of a 1:8 scale model test of a cylindrical containment indicated a failure pressure of four times the design pressure (NUREG/CR-4209) and that of a 1:6 scale model of a cylindrical reinforced concrete containment indicated a failure pressure of three times the design pressure (NUREG/CR-5476). However, these predictions are all based on the assumption of an undegraded containment. NRC currently does not have any insights or information on the performance of a degraded containment under severe accident conditions.

Resolution: In a memorandum from W.T. Russell to D.L. Morrison, " User Needs for Degraded Containment Research," dated 05/08/96, NRR requested RES to perform research on the effects grease leakage through the containment concrete from tendon sheaths has on containment structural integrity of PCC. The objective of the proposed research is to determine the effect of leaking grease on the structure capacities of PCCs under accident conditions up to and including severe accidents. The research is to identify whether or not grease leakage leads to a loss of containment strength such that the ability of a containment with grease leakage to meet the design basis is brought into question, or if non-negligible increase in risk occur as a result of reductions in ultimate containment capacity. If this concern is confirmed, NRR will then propose that the industry address the issue.

Completion Date: 06/17/96 0

06/30/96 A.E-46 NUREG-0933

_ _ _ ~

i GCCA-0059: GENERIC BWR STRAINER CLOGGING TAC No.: M86925

Contact:

J.W. Shapaker Regr_iDtion: 10 CFR 50.46 requires that adequate ECCS flow be provided to maintain the core temperature at an acceptably low value and to remove decay heat for the extended period of time required by the long-lived radioactivity remaining in the core following a DBA. Therefore, based on operating experiences at several domestic and foreign reactors, the NRC issued Bulletin 96-03 to request that holders of operating licenses for BWRs implement appropriate procedural measures and plant modifications to ensure the capability of the ECCS to perform its safety function following a LOCA. The NRC identified three potential resolution options; however, a licensee may propose an alternative approach that provides an equivalent level of assurance that the ECCS will be able to perform its safety function following a LOCA.

Oriainatina Document: NUREG/CR-6224, " Parametric Study of the Potential for BWR ECCS Strainer Blockage Due to LOCA Generated Debris," published October 1995.

Reaulatory Assessment: The actions requested by Bulletin 96-03 are considered backfits under the terms of 10 CFR 50.109(a)(4)(i) and are necessary to ensure that licensees are in compliance with existing NRC rules and regulations.

Nevertheless, the resolution approach presented in the bulletin provides an interpretation of what licensees are expected to do to comply with 10 CFR 50.46 that heretofore has not been documented as an NRC position for the nuclear power O

industry.

Resolution: Issuance of Bulletin 96-03, " Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors," dated 05/06/96 (Accession No. 9605020119).

Comoletion Date: 05/06/96 GCCA-0060: INADEOUATE TESTING OF SAFETY-RELATED LOGIC CIRCUITS TAC No.: M90863

Contact:

J.W. Shapaker

==

Description:==

The NRC has issued numerous ins regarding problems with testing of safety-related logic circuits: IN 88-83, " Inadequate Testing of Relay Contacts in Safety-Related Logic Circuits," dated 10/19/88; IN 91-13, " Inadequate Testing of Emergency Diesel Generators (EDGs)," dated 03/04/91; IN 92-40, " Inadequate Testing of Emergency Bus Undervoltage Logic Circuitry," dated 05/27/92; IN 93-15,

" Failure to Verify the Continuity of Shunt Trip Attachment Contacts in Manual Safety Injection and Reactor Trip Switches," dated 02/18/93; and IN 93-38,

" Inadequate Testing of Engineered Safety Features Actuation Systems," dated 05/24/93. Despite these notices, events occurred similar to those described in the ins which indicated that licensees had not taken sufficient action to correct previously identified problems in logic circuit surveillance testing. On 03/07/95, NRC issued IN 95-15, " Inadequate Logic Testing of Safety-Related Circuits," which informed licensees about the more recent events. Nevertheless, n

the NRC determined that licensees should review their surveillance procedures for V

the RPS, EDG load-shedding and sequencing, and actuation logic for the ESFS to ensure that complete testing is being performed as required by the TS.

06/30/96 A.E-47 NUREG-0933

Oriainatina Document: Multiple events at nuclear power reactors.

Reaulatory Assessment: A number of NRC regulations document the requirements to test safety-related systems to ensure that they will function as designed when called upon.10 CFR 50.36, Paragraph (c)(3), " Technical Specifications," states that, " surveillance requirements are requirements relating to test, calibration or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within the safety limits, and that the limiting conditions of operation will be met." Surveillance requirements to assure continued operability of safety-related logic circuits have been included in the plant-specific TS for all operating nuclear power plants. Other documents that provide a basis for these requirements include:

10 CFR 50.55a, " Codes and Standards," paragraph (h) which includes reference to IEEE Standard 279, " Criteria for Protection Systems for Nuclear Power Generating Stations";

Appendix A to 10 CFR 50, GDC 21, " Protection System for Reliability and Testability";

Appendix A to 10 CFR 50, GDC 18, " Inspection and Testing of Electric Power Systems";

Appendix B to 10 CFR 50, Criterion XI, " Test Control";

RG 1.118, " Periodic Testing of Electric Power and Protection Systems";

RG 1.32, " Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants."

Therefore, the actions requested by GL 96-01 are considered backfits under the terms of 10 CFR 50.109(a)(4)(i) and are necessary to ensure that licensees are in compliance with existing NRC rules and regulations.

Resolution: Issuance of GL 96-01, " Testing of Safety-Related Logic Circuits,"

dated 01/10/96 (Accession No. 9601050193).

Comoletion Date: 02/27/96 GCCA-0061: BORAFLEX DEGRADATION IN SPENT FUEL POOL STORAGE RACKS TAC No.: M91447

Contact:

J.W. Shapaker Descriotion: The degradation of Boraflex that has been observed in spent fuel storage racks has been addressed by the NRC in several ins: 87-43 (Accession No.

8709010085); 93-70 (Accession No. 9309070206), and 95-38 (Accession No.

9509050009). Furthermore, EPRI has been studying the phenomenon of Boraflex degradation for several years and has identified two concerns with respect to using Boraflex in spent fuel storage racks. The first is related to gamma radiation-induced shrinkage of Boraflex and the potential to develop tears or gaps in the material. This aspect is typically accounted for in criticality analyses of spent fuel storage racks. The second concern is the long-term 06/30/96 A.E-48 NUREG-0933

Boraflex performance throughout the intended service life of the racks as a result of gamma irradiation and exposure to the wet pool environment.

Oriainatina Document: Numerous reports of Boraflex degradation.

Reaulatory Assessment: On the basis of test and surveillance information from plants that have detected areas of Boraflex degradation, no safety concern exists that warrants immediate action. Boraflex dissolution appears to be a gradual and localized effect, forewarned by relatively high silica levels in the pool water.

This occurrence of increased pool silica is more pronounced in PWRs than BWRs because of the greater effectiveness of silica removal by the BWR demineralizers in the non-borated pool water environment. Because of the safety margin present in spent fuel storage pools, compliance with the required sub-criticality margin (or conformance with the same margin to which licensees have committed in their updated FSARs) can be expected to be maintained during the initial stage of Boraflex degradation. This safety margin is due to the conservatism in treating the reactivity effects of possible variations in material characteristics and mechanical tolerances and the generally lower reactivity of stored fuel than that assumed in the safety analysis. However, to verify compliance with both the regulatory requirements of GDC 62 and the 5% sub-criticality margins, either contained in the TS or committed to in the updated FSARs, and to maintain an appropriate degree of defense-in-depth measures, the staff has concluded that it is appropriate for licensees to submit information under the provisions of Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR Part 50.54(f).

O()

Resolution: Issuance of GL 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks," dated 06/26/96 [ Accession No. 9606240132].

Completion Date: 06/26/96 GCCA-0062: ANSYS AND GTSTRUDL COMPUTER PROGRAM ERROR NOTIFICATIONS TAC No.: M91542

Contact:

E.Y. Wang Descriotion: Boeing Computer Services (BCS) submitted a 10 CFR 21 notification describing errors identified in ANSYS and GTSTRUDL computer programs. Both programs are used for the design and analysis of safety-related nuclear power plant structures, systems and components, and more recently, for the design and analysis of plant modifications. There were five organizations that BCS was unable to contact because the organizations have moved or dissolved. Mail to these five organizations was returned to BCS by the US Postal Service.

Oriainatina Document: Part 21 Log Number 95-078, a BCS Error Notification, dated 02/13/95 (Accession No. 9502220192).

Reaulatory Assessment: Neither NRC nor BCS was able to evaluate the safety significance of some of the errors. Some of these' errors that can prevent execution of the program sub-routines do not impact the safety-related calculations. In addition, utilities and other organizations that have received O

notification from BCS reported that the errors identified did not affect any safety-related calculations.

06/30/96 A.E-49 NUREG-0933

Resolution: IN 96-29, " Requirements in 10 CFR Part 21 For Reporting And Evaluating Software Errors," was issued on 05/20/96 to alert the industry of the potential problem (see Accession No. 9605150209).

Comoletion Date: 05/20/96 GCCA-0063: INADE00 ATE CONTROL OF MOLDED-CASE CIRCUIT BREAKERS TAC No.: M91622

Contact:

J.R. Tappert

==

Description:==

Exercising of circuit breakers prior to surveillance testing does not provide good information on the as-found condition of the breakers. During inspections, both Diablo Canyon and South Texas licensees were found to cycle their circuit breakers prior to performing overcurrent surveillance testing.

Periodic inspection and testing of circuit breakers in their as-found condition is an appropriate way of demonstrating the functional operability of the breaker and of detecting any degradation.

Oriainatina Documents: Inspection Reports 50-275, 323/94-27, dated 12/21/94 (Accession No. 9501060003), and 50-498, 499/94-35, dated 01/19/95 (Accession No.

9502010083).

Reaulatory Assessment: The practice of pre-conditioning before testing (e.g.,

lubricating pivot points and manually cycling the breaker) defeats the purpose of the periodic test. Such pre-conditioning does not confirm continued operability between tests, nor does it provide information on the condition of the circuit breaker for trending purposes. Testing some circuit breakers in the as-found condition can provide useful data on which to base decisions on surveillance intervals and the ability of the untested circuit breakers to perform their intended function. Since only a fraction of the circuit breakers are tested each refueling outage to justify the operability of the remaining circuit breakers, pre-conditioning before testing does not provide the expected assurance of the operability of remaining breakers which are not tested. By pre-conditioning circuit breakers, useful information may be lost because the breaker may not have been capable of performing its intended function without pre-conditioning.

Resolution: IN 96-24, " Preconditioning of Molded-Case Circuit Breakers before Surveillance Testing," was issued on 04/25/96 (Accession No. 9604220229).

Comoletion Date: 04/25/96 GCCA-0064: RELOCATION OF RCS PRESSURE / TEMPERATURE UMITS TAC No.: M91749

Contact:

J.W. Shapaker

==

Description:==

During the development of the improved STS, a change was proposed to relocate the pressure / temperature (P/T) limit curves and LT0P setpoint curves and values currently contained in the TS to a licensee-controlled document. As one of the improvements to the STS, the staff agreed with the industry that the curves and setpoints may be relocated outside the TS to a licensee-controlled document so that the licensee could maintain these limits efficiently and at a 06/30/96 A.E-50 NUREG-0933

O) lower cost, provided that the parameters for constructing the curves and y

setpoints are derived using a methodology approved by the NRC.

Oriainatina Document: Improved STS.

Reaulatory Assessment: Any action by licensees to propose changes to TS in accordance with the guidance in GL 96-03 is voluntary and, therefore, is not a backfit under 10 CFR 50.109.

Resolution: GL 96-03, " Relocation of the Pressure / Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," was issued on 01/31/96 (Accession No. 9601290350).

Comoletion Date: 01/31/96 GCCA-0065: RECONSIDERATION OF PLANT SECURITY RE0VIREMENTS TAC No.: M91896

Contact:

J.W. Shapaker

==

Description:==

In an SRM dated 02/18/94, the Commission endorsed staff recommendations to: (1) issue generic correspondence informing licensees of the opportunity to revise certain commitments in their security plan; and (2) proceed with rulemaking regarding specific changes to reduce or eliminate certain security requirements. GL 96-02 identifies those areas in which licensees may choose to revise their plans without having to wait for the issuance of rule O

changes.

V Oriainatina Document: SRM dated 02/18/94.

Reaulatory Assessment: The NRC has reconsidered its positions on certain security measures associated with protecting nuclear power plants against an internal threat. Suggestions contained in the GL for the reduction or elimination of security requirements that provide only marginal protection against the insider threat are not NRC requirements, and no specific action or written response is required. Some of the suggested security plan changes require the submittal of a license amendment, in accordance with 10 CFR 50.90, while other changes may be processed in accordance with the provisions of 10.CFR 50.54(p) and can be implemented without NRC approval.

Resolution:

GL 96-02,

" Reconsideration of Nuclear Power Plant Security Requirements Associated With an Internal Threat," was issued on 02/13/96 (Accession No. 9601230206).

Completion Date: 02/13/96 GCCA-0066: FIRES IN EMERGENCY DIESEL GENERATOR EXCITEM TAC No.: M92594

Contact:

J.R. Tappert p

DescriottgII: Potential for a fire in an emergency diesel exciter during operation Q

following undetected fuse blowing. On 09/30/94 at Wolf Creek, the "A" train EDG main power potential transformer of the static cxciter-voltage regulator caught 06/30/96 A.E-51 NUREG-0933

fire after about one hour of sustained high power operation. On 10/11/94, the "B" EDG exciter caught fire under similar circumstances. After each fire, the licensee found that one of the 100-amp fuses in the secondary circuits of the exciter potential transformer had blown. It was subsequently determined that the fuses had blown due to manual engine shutdown without exciter shutdown. Because there was no blown-fuse indication, the normal full and above full-power runs for routine testing were conducted subsequently without knowing that the fuses had blown and " single phased" the potential transformers. The " single phased" potential transformers became overloaded, suffered progressive insulation breakdown, and eventually caught fire.

Oriainatina Document: Inspection Report 50-482/94-13, dated 11/16/94 (Accession No. 9411220024).

Reaulatory Assessment: The licensee has installed blown fuse indications on the EDG exciter cabinets. The licensee has also installed volts-per-hertz protection to avoid this potential failure mode. In other designs, under-frequency protection is often available that will independently shut down the exciter upon loss of the prime mover. However, EDG exciter systems that remain on, either through system design flaw or malfunction, after engine mechanical shutdown may fail in a manner similar to that experienced by Wolf Creek. An IN was issued to alert licansees to this concern.

Resolution: IN 96-23, " Fires in Emergency Diesel Generator Exciters During Operation Following Undetected Fuse Blowing," was issued on 04/22/96 (Accession No. 9604170169).

Comoletion Date: 04/22/96 GCCA-0067: INADE0VATE CAPACITY OF CCW LEADS TO FREON RELEASE TO THE CONTROL ROOM TAC No.: M92595

Contact:

W.F. Burton Descriotion: On 11/14/94, the licensee for Fort Calhoun initiated a plant shutdown because an engineering analysis had shown that the control room air-conditioners could be disabled by a large primary coolant system pipe rupture or a main steamline break inside the containment. This could result in creating an environment in the control room which could hinder operator activities and increase temperatures above the design temperatures of safety-related equipment in the control room.

Oriainatina Document: Event Notification 28029, dated 11/14/94.

Reaulatory Assessment: A large primary coolant system pipe rupture or main steamline break inside the containment could cause the closed-cycle cooling water (CCW) temperature to rise rapidly because of the large heat input from the containment coolers during these postulated accidents. As a result, the CCW temperature could exceed the maximum post-accident CCW temperature specified in the FSAR, as well as the temperature used to calculate thermal stresses in certain piping segments.

The control room air-conditicning units are equipped with rupture discs that are designed to blow out on high CCW temperature. If the refrigerant was released, 06/30/96 A.E-52 NUREG-0933

f O

with the air-conditioning units could not be recovered. Without air-conditioning, and the control room ventilation system operating in the emergency pressurization mode, the control room temperature could increase to levels that could hinder operator activities and cause the design temperatures of safety-related equipment in the control cabinets to be exceeded.

The complex nature of CCW systems may prevent coirect identification of the most limiting potential operating configuration of the system. Certain safety-related components served by CCW systems, such as air-conditioning units and EDGs may fail in a non-recoverable manner as a result of these temperature transients. As a result of a loss of CCW, safety-related systems which depend on CCW for cooling may become inoperable. An IN has been issued to alert licensees of the need to identify the most limiting system configuration and provide the proper procurement specifications for the air-conditioning units.

Resolution: IN 96-01, " Potential for High Post-Accident Closed-Cycle Cooling Water Temperatures to Disable Equipment Important To Safety," was issued on 01/03/96 (Accession No. 9512270372).

Comoletion Date: 01/03/96 GCCA-0068: BWR STABILITY WITH FLOW SLIGHTLY LESS THAN NATURAL CIRCULATION FLOW TAC No.: M92601

Contact:

T.J. Carter O

Descriotion: Two instances of power operation have occurred in which the core flow, following run-back of the recirculation pumps, appeared to be less than that normally attributed to natural convection flow as shown on the power / flow maps available to reactor operators. Since this does not fit the expected response, the operators may not understand what could be happening to cause this observation.

Oriainatina Document: Morning Report 1-95-0078, dated 06/06/95.

Reaulatory Assessnie_nl: An apparent flow less than that associated with natural circulation may be ex)lained by a number of subtle things. The power / flow maps are general curves anc should not be assumed to present precise values, and flow instrumentation is not calibrated for flow rates near that resulting from natural convection. For these reasons, the staff does not believe there is a significant safety concern. An IN should be issued to inform the operators of the various mitigating considerations thereby clarifying their response to such an observation.

Resolution: IN 96-16, "BWR Operation With Indicated Flow Less Than Natural Circulation," was issued on 03/14/96 [ Accession No. 9603110159).

Comoletion Date: 03/19/96 k

06/30/96 A.E-53 NUREG-0933

}

GCCF 0069: TERRY TURBINE DEPENDABILITY TAC No.: M92636

Contact:

T.J. Carter

==

Description:==

Over the years there have been a number of operability problems associated with Terry turbines. The turbine's principal use is as the driver for pumps in the AFW system at PWRs and in the RCIC system at BWRs, both of which are safety-related systems. Since not every licensee experiences the same problems, there has not been a universally endorsed effort to improve the operability and reliability of turbine-driven pumps. As a result, evaluations have been made of individual problems after they occur, and some facilities have implemented corrective actions. No overall assessment that considers all the known problems with universal implementation has occurred.

Oriainatina Document: NUREG-1275, Vol. 10, "AE00 Operating Experience Feedback Report - Reliability of Safety-Related Steam Turbine-Driven Standby Pumps,"

October 1994, and numerous recently reported events.

Reaulatory Assessmet: The sequences where the turbine-driven pumps are most important to safety are those that involve a loss of all AC power, or during an SB0, particularly at PWRs. At BWRs, there are diverse or redundant means of coping with an SB0 and the safety significance of a failure of a single Terry turbine is reduced. For SB0 scenarios, the amount of decrease in the conditional damage frequency that can be gained by increasing the reliability of the turbine-driven AFW pump is limited by risk associated with a seal-LOCA. The perceived unreliability or unavailability of these machines is often overstated, given the fact that essentially all failures of concern are recoverable within a short period of time and, therefore, the pumps can still be available for mitigating SB0 scenarios. Thus, there is no real immediate safety concern that requires immediate regulatory response or action by the staff.

Resolution: Memorandum from L.B.

Marsh to G.M.

Holahan,

" Terry Turbine Reliability and Future NRC Actions (TAC No. M92636)," dated 05/15/96. The conclusion was that existing and proposed rules are available to obtain needed information and achieve adequate system reliability including remedial actions.

Industry should be allowed to resolve the issue through their own initiatives.

Comoletion Date: 06/15/96 GCCA-0070: EVALUATE IMPACT OF RCP SUPPORT COLUMN TILT ON LEAK BEFORE BREAK ANALYSES TAC No.: M93024

Contact:

C.V. Hodge Descriotion: To avert interference with crossover leg piping, the support designer at a PWR plant arranged for the base of the front inside support column for a RCP to be placed 6 to 12 inches closer to the reactor pressure vessel than in the original design. W, the NSSS vendor, determined that the resulting 2 to 5 degree tilt in the column decreased the vertical stiffness by a small amount.

More significantly, however, W determined that it might degrade the thermal expansion and loop loadings of the system. W identified the affected plants:

Callaway, Seabrook, Watts Bar-1 & 2, Comanche Peak-1 & 2, Sequoyah-1 & 2, Wolf Creek, Farley-1 & 2, Summer, Vogtle-1 & E, and Shearon Harris.

06/30/96 A.E-54 NUREG-0933

Oriainatino Document: Part 21 Log No. 95016, " Closeout of an Interim Report of an Evaluation of a Deviation or Failure to Comply Pursuant to 10 CFR 21.21(a)(2)," dated 11/15/94 (Accession No. 9411280210).

Reaulatory Assessment: The safety significance of this problem is the potential decrease in analyzed safety margin against unstable crack growth in an RCS loop.

The maintenance of the safety margin is needed to justify the application of leak-before-break criteria.

Resolution: EMCB reviewed the Part 21 notification for leak-before-break (LBB) analysis, determined a need for additional information from licensees of the affected plants, and requested that project managers for the affected plants request the following data: (1) comparison of applied loads used in the initial LBB analysis and the increased loads as a result of column tilt; (2) comparison of margins on critical crack size to leakage crack size with and without the column tilt; (3) crack stability analysis using the increased piping loads; and (4) demonstration that the leakage crack will not. experience unstable crack growth even if larger loads (at least sqrt(2) times the normal plus safe shutdown earthquake loads) are applied. The factor sqrt(2) may be reduced to 1.0 if the loads are combined absolutely. On the basis of information obtained. EMCB determined that all affected licensees have reviewed and responded to the H Nuclear Safety Advisory Letter NSAL-94-025, dated 11/10/94, (Accession No.

9607260276), on this subject. All licensees reported that the primary system piping in their plants satisfy the necessary margins in the LLB analyses.

Comoletion Date: 06/04/96 GCCA-0071: FISH MOUTH BURST AND BOWING OF PREVIOUSLY-PLUGGED STEAM GENERATOR TUBES TAC No.: M93227

Contact:

E.J. Benner Descriotion: During an SG inspection at Haddam Neck, a " fishmouth" opening in a plugged tube was observed. In addition, the tube was reported to be bowed towards the tube lane by 0.5 to 0.75 inches.

Oriainatina Document: NRC Morning Report I-95-0034, dated 03/01/95.

Reaulatory Assessment: H has previously analyzed burst plugged tube conditions because of previous events. The reported bowing of tubes is a new development in that contact of the adjacent tubes is possible. An IN has been proposed to inform licensees of H recommendations for burst and bowed tubes that leaking plugs be replaced, if leakage is detected during future outages, and that tubes which are adjacent to plugged tubes be inspected for signs of wear, at all future outages, until remedial action has been taken to remove the potential for plugged tubes to be pressurized to burst.

The safety significance of this concern is low because calculations of dynamic interaction of the burst tube with neighboring tubes indicate that significant wear of the neighboring tubes would not be expected in the course of one cycle of operation.

06/30/96 A.E-55 NUREG-0933

Resolution: Because of perceived low safety significance of the concern, the NRR/AE0D/RES Events Assessment Panel cancelled development of the IN with the agreement of DE on 04/02/96.

Completion Date: 03/08/96 GCCA-0072: BLOCKAGE OF UNTESTED ECCS PIPING TAC No.: M93360

Contact:

E.J. Benner

==

Description:==

During a refueling outage, it was discovered that two out of the four ECCS lines from the containment sump were partially blocked by debris buildup. It is presumed that the debris was present from initial construction.

Oriainatino Document: International Reporting System Report 1505.G0 (Proprietary) dated 07/05/95.

Reaulatory Assessment: Investigations by the Spanish regulatory agency revealed that several segments of ECCS piping are not subject to periodic functional flow-testing because they are not used during operation. The potential exists for undetected blockage of these segments of pipe. This concern is potentially generic. The safety significance of the concern is that debris in this piping can degrade system performance by reducing system flow rates and/or damaging valves, pumps, and heat exchangers.

Previous NRC generic communications (Bulletin 93-02 and Supplement 1, " Debris Plugging of Emergency Core Cooling Suction Strainers," dated 05/11/93 and 02/18/94, Accession Nos. 9305110015 and 9402180174), as well as several ins have addressed aspects of this concern. Thus, an IN is the appropriate generic action to provide licensees with the additional information provided by the foreign experience.

Resolution: IN 96-10, " Potential Blockage by Debris of Safety System Piping Which Is Not Used During Normal Operation or Tested During Surve111ances," was issued on 02/13/96 (Accession No. 9609070259).

Comoletion Date: 02/13/96 GCCA-0073: PORV INOPERABILITY MASKED BY DOWNSTREAM INDICATIONS DURING TESTING TAC No.: M93400

Contact:

E.J. Benner

==

Description:==

On 08/09/95, surveillance testing of the PORVs at St. Lucie indicated that they were not operating properly. The PORVs

' e removed and checked on a test bench. Both valves failed to relieve at any

tuating delta pressure across the main valve disk to pilot vent path. The PORVs at St. Lucie are credited during feed-and-bleed and LTOP scenarios and can be used for pressure control to limit opening of SRVs.

Oriainatina Document: Event Notification 29178, dated 08/10/95.

06/30/96 A.E-56 NUREG-0933

O)

Reaulatorv Assessment: The licensee discovered that the valve guide bushings were

\\

installed backwards. The bushing has holes at one end to allow pressure beneath the main valve disk to be vented when the pilot valve is actuated, allowing the valve to operate. With the bushing installed backwards, there was no vent path from the main valve to the pilot line. This condition probably existed since the last refueling outage (approximately 10 months ago) when a contractor, using approved licensee procedures, worked on the valves. The licensee has subsequently installed the bushings correctly, and the PORVs have tested satisfactorily.

A potential generic concern exists with the use of tailpipe acoustic monitors to verify operation of PORVs. Since the valves were refurbished, surveillance testing was performed twice with results inoicating satisfactory performance of the PORVs, as evidenced by tailpipe acoustic data. The licensee subsequently deterrined that internal clearances in the main valve are sufficient to pass medi ibrough the pilot valve, providing positive indication on the acoustic monii

., despite the failure of the main valve to lift. There has been no previw evidence that PORVs have remained inoperable for extended periods of time because of inadequate surveillance testing. The currently proposed generic action is to issue an IN alerting licensees to the incorrectly installed bushings and the potential inadequacy of acoustic monitors to verify valve operation. An IN is the appropriate generic action to inform licensees of potential limitations of the use of acoustic monitors to verify operation.

Resolution: IN 96-02, "Ir. operability of Power-0perated Relief Valves Masked by Downstream Indications During Testing," was issued on 01/05/96 (Accession No.

9512290129).

Comoletion Date: 01/05/96 GCCA-0074: LOSS OF RC INVENTORY AND POTENTIAL LOSS OF EMERGENCY NITIGATION FUNCTIONS WHILE IN A SHUTDOWN CONDITION TAC No.: M93568

Contact:

E.N. Fields

==

Description:==

On 09/17/94 at Wolf Creek, operators were attempting to re-borate RHR Train B while, at the same time, maintenance personnel were repatking an RHR Train A to Train B crossover isolation valve. Train B is re-borated by rec.irculating water through a loop that contains the RHR system piping, the refueling water storage tank (RWST), a containment spray pump, a manual RWS1 isolation valve, and a RHR system crossover line.

When the RWST isolation valve was opened for the re-boration process and the Irain A to Train R crossover isolation valve was opened for s^roke timc testing, a drain-down path was inadvertently created from the RCS to the RWST. As a result, an unintentional RCS flow path was created allowing approximately 35,000 liters (9,200 gallons) of reactor coolant to transfer to the RWST.

Oriainatina Document: Region IV Horning Report Number 4-94-0100.

Reaulatory Assessment: If the drain-down had not been promptly tarminated, the

,p operability of the ECCS would have been compromised. Also, RCS water flashing to

\\

steam in the piping or in the RWST would likely have created conditions conducive to water hammer. This event presented an immediate safety concern.

06/30/96 A.E-57 NUREG-0933

Resolution: A Ta3k Action Plan was developed to inform licensees and the Regions of this vulnerability, to request all licensees to take appropriate measures to prevent a similar event, and to implement a long-term resolution. As part of this Task Action Plan, IN 95-03, " Loss of Reactor Coolant Inventory and Potential Loss of Emergency Mitigation Functions While in a Shutdown Condition," was issued on 01/18/95 (Accession No. 9501110412). This IN was updated by IN 95-03, Supplement 1 (Accession No. 9602050208) of the same title. The Supplement was developed under TAC No. M93568 and describes further staff insights into this event.

Completion Date: 03/25/16 GCCA-0075: CONTROL ROD DRIVE MECHANISM PENETRATION CRACKING TAC No.: M93641

Contact:

E.N. Fields

==

Description:==

The activity of CRD penetration cracking has been ongoing since an incident at a foreign reactor in 1991 identified cracks in the control rod drive mechanism (CRDM) penetrations. Three U.S. licenseer have voluntarily conducted inspections of CRDM penetrations, two of which have identified cracks.

Approximately 20 cracks in one penetration were identified at Oconee and three cracks were identified in one penetration at D.C.

Cook. A foreign plant experienced several demineralizer resin bed intrusions which were concluded to have resulted in extensive CRDM punetration cracking. The IN will notify the industry that resin intrusicas may result in accelerated corrosion of CRDM penetrations and of other components fabricated from Alloy 600.

Oriainatina Document: Memorandum from R. Herman to A. Chaffee, dated 09/01/95.

Reaulatory Assessment: Thare does not appear to be an insediate safety concern.

No CRDM penetration failures have been experienced at U.S. reactors to date. The staff is net aware of any significant primary system resin bed intrusion at any U.S. PWR.

Rg;,olution: IN 96-11, " Ingress i Demineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations," was issued on 02/14/96 (Accession No. 9602090038).

Completion Date: 02/14/96 GCCA-0076: AUGMENTED EXAMINATION OF REACTOR VESSEL TAC No.: M93643 Cont aI;_t,: E.J. Benner Cescription: 10 CFR 50.55a (g)(6)(ii)(A), " Augmented Examination of Reactor Vessel," was issued in 1992 to require a one-time augmented inspection of the reactor vessel in accordance with the 1989 Edition of Section XI of the ASME Code. Several licensras were unaware of the rule or had misconceptions regarding staff approvals required by the rule when complete inspection coverage cannot be achieved.

Oriainatina Document: Issuance of new rule 10 CFR 50.55a(g)(6)(ii)(A). Federal Register Notice 57 FR 34673, dated 08/06/92.

06/30/96 A.E-58 NUREG-0933

l I

l{(7 Reaulatory Assessment: TAC No. M93643 was issued to DE to develop an IN to alert

,('j licensees to the presence of the rule and to provide clarification of the rule to licensees after it became apparent that several licensees were unaware of th-i rule or had misconceptions regarding staff approvals required by the rule when complete inspection coverage cannot be achieved (see TAC No. M67462).

An IN is an appropriate generic resolution. Greater generic action is not warranted because of the pre-existing presence of the rule. The IN is warranted because of evidence that licensees are either unaware of the rule or unaware of the full ramifications of the rule.

Resolution: IN 96-32, " Implementation of 10 CFR 50.55a(g)(6)(ii)(A), ' Augmented Examination of Reactor Vessel,'" was issued on 06/05/96 (Accession No.

9605200277).

Comoletion Date: 06/05/96 GCCA-0077: CLOSED HEAD VENT CAUSES INACCURATE LEVEL INDICATION DURING REDUCED INVENTORY IAC No.: M93751

Contact:

R.A. Benedict

==

Description:==

Closed head vent causes inaccurate level indication during reduced inventory. On 09/13/95, Surry-1 was in a shutdown condition, cooled down and depressurized preparatory to refueling. A pressurizer PORV and its block valve h

had been opened, connecting the pressurizer to the pressurizer relief tank (PRT)

C/

which was pressurized with nitrogen to 11 psig. The reactor head vent was open to the top of the water-level indicating standpipe through the vent connection to the PORV relief line. The water level in the reactor and pressurizer had been lowered to slightly below the reactor vessel flange so that the vessel head studs could be de-tensioned. The pressurizer was empty and the reactor coolant piping was full to part way up the surge line.

In order to install the cavity seal ring so that the cavity could be flooded up to permit lifting the reactor head, the head vent valve was closed and the ventline spoolpiece was disconnected. After the seal ring was in place, the spoolpiece was re-connected but the head vent valve was not reopened as it should have been. This resulted in loss of function of the only reactor water level indication available while the reactor head was still installed.

Letdown and makeup were being maintained manually by an operator who monitored the standpipe level. As pressurizer relief tank overpressure was being reduced, the operator saw an increase in indicated level due to water in the reactor vessel being forced up into the standpipe and surge line as the gas bubble trapped in the reactor vessel expanded. The operator increased letdown to maintain the indicated level required. This process continued periodically for about five hours, resulting in reactor coolant inventory baing reduced by almost 5,000 gallons.

When the relief tank pressure was subsequently reduced to atmospheric pressure

^

and the reactor vessel studs were de-tensioned, the vessel head lifted enough to

,(m) relieve the gas pressure in the head. This caused a sudden drop in indicated i

06/30/96 A.E-59 NUREG-0933

i level in the standpipe by about five feet. The operator immediately took action to restore the level to where it was supposed to be.

i Oriainatina Document: NRC Region II Morning Report Number 2-95-0083, dated l

09/01/95.

l l

Reaulatory Assessment: The safety significance of this particular event is l

minimal: the water level remained 1.5 feet above mid-loop and more than six feet above the core. Forced circulation residual heat removal continued. If the gas l

bubble had continued to expand, the water level would not have dropped lower than just below the top of the reactor coolant hot-leg pipe; at that point, the gas bubble would vent through the hot-leg pipe into the pressurizer and into the pressurizer relief tank, equalizing pressure between the relief tank and the vessel atmosphere. The standpipe indicated level would then have been correct.

Previous events have occurred in which gas bubbles and level indication problems have been of concern. These events have been the subjects of previously-issued ins. An IN is the appropriate generic action to inform licensees of considerations involved in this present Surry event.

Resolution: IN 96-37, " Inaccurate Reactor Water Level Indication and Inadvertent Draindown During Shutdown," was issued on 06/18/96 (Accession No. 9606120154).

Comoletion Oa,13: 06/18/96 GCCA-0078: SHUTDOWN COOLING FLOW BYPASSING CORE RESULTS IN TEMPERATURE AND PRESSURE INCREASES TAC No.: M93752

Contact:

C.V. Hodge D_escriotion: On 10/03/95, the Events Assessment Panel classified the set of two undetected mode changes of the BWR at Hope Creek as a Significant Event for the NRC Performance Indicator Program. The basis for this classification is the failure of the Hope Creek licensee to comprehend the condition of the plant for an extended period of time.

Oriainatina Document: Preliminary Notification of Event or Unusual Occurrence, PNO-I-95-026, " Improper Reactor Recirculation System Alignment Led to Shutdown Cooling Flow Bypassing the Reactor Core," dated 08/02/95.

Reaulatory Assessment: On 08/09/95, the licensee reported that a shutdown cooling bypass event had occurred at 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> on 07/08/95, when the operating crew left the "B"

recirculation loop discharge valve in a partially open position to i

mitigate potential thermal binding of that valve. During the shutdown cooling l

evolution, 2,000 gpm of RHR flow was diverted through the open recirculation i

valve and redirected from the intended path (through the core) to a parallel path l

(through recirculation Loop "B"). This parallel path bypassed the core. In i

addition, operators later secured the RHR system, in accordance with plant I

procedures, to facilitate testing of the shutdown cooling isolation valves. The l

2,000 gpm bypass flow combined with the isolation of the RHR system resulted in l

the heatup of the RCS that caused the first undetected mode change. Shutdown cooling was then returned to service. About ten hours later, bypass flow i

I increased to approximately 4,000 gpm when the recirculation valve was further 1

06/30/96 A.E-60 NUREG-0933

opened in an attempt to re-close it, causing the second undetected mode change.

The valve was manually closed at 0550 hours0.00637 days <br />0.153 hours <br />9.093915e-4 weeks <br />2.09275e-4 months <br /> on 07/09/95, terminating the event.

Licensee investigation into this event identified key corrective actions in the areas of operator training, operator procedure compliance, valve thermal binding assessment, and management response to the event. The licensee determined on 08/04/95 that an operational condition change occurred from cold shutdown to hot shutdown (the undetected first mode change); this was not known at the time of the event. As a result of both these unplanned mode changes, several TS LCOs were not met.

The NRC conducted a special inspection of circumstances surrounding this event and concluded that this event was initiated when plant operators inappropriately left open the recirculation pump discharge valves, allowing shutdown cooling flow to bypass the reactor vessel, which decreased the ability of the shutdown cooling system to remove decay heat and allowed the RCS temperature and pressure to increase. This resulted in an undetected change in the plant operational condition from the desired cold shutdown to the hot shutdown condition (the undetected first mode change).

Resolution: The inspection team concluded that the principal causes of this event were inadequate communications and failure to follow procedures and that contributing causes were poor quality procedure instructions and inadequate training. In addition, senior plant management initially failed to correctly assess the significance of this event. The failure resulted in a 10-day delay in initiating a comprehensive root cause evaluation and contributed to the failure to make the required notification to the NRC. The team also concluded that this O-event was safety significant. However, the consequences of the event w:.re minimal and the event had no direct adverse effect on the health and safety of the public or plant personnel. The identified weakness in both operator and management performance during and following this event were also significant. On 09/13/95, the Events Analysis and Generic Communications Branch (PECBB) briefed senior NRC management on the event and, on 10/05/95, PECBB and Region I briefed the ACRS.

On 01/18/96, IN 96-05, " Partial Bypass of Shutdown Cooling Flow from the Reactor Vessel," was issued describing this event.

Comoletion Date: 01/19/96 GCCA-0079: POTENTIAL CONTAINMENT LEAK PATH THROUGH HYDROGEN ANALYZER TAC No.: M93753

Contact:

J.R. Tappert

==

Description:==

Hydrogen analyzers communicate with the containment atmosphere after an accident and can create a containment leak path. Two plants recently identified potential containment leak paths through the hydrogen monitor system.

At Braidwood, a potential containment bypass existed for several months when a hydrogen sensing line was not re-connected following an integrated containment leak rate test. At Catawba, the hydrogen analyzer was tested in a de-energized condition. After an accident, the analyzer would be energized and the pressure boundary changes. The testing would not reveal any leakage in the energized pressure boundary.

Oriainatino Document: LER 50-457/95-02-01, dated 04/21/95 (Accession No.

9504240323).

06/30/96 A.E-61 NUREG-0933

i Reaulatory Assessment: Because containment penetrations, systems, and equipment that will be exposed to the containment atmosphere must be leak rate-tested to ensure that containment integrity is maintained after a DBA, the procedures for l

these tests must adequately consider the penetration configuration. Additionally, l

because hydrogen monitor containment isolation valves are normally procedurally l

opened after a DBA, any leakage in the hydrogen monitor system may bypass the l

containment and can challenge regulatory radiological guidelines. An IN was l

issued to alert licensees to these concerns.

Resolution: IN 96-13, " Potential Containment Leak Paths Through Hydrogen Analyzers," was issued on 02/26/96 (Accession No. 9602220234).

Completion Date: 02/26/96 GCCA-0080: INADE0VATE TESTING AND DESIGN OF TORNADO DAMPERS TAC No.: M93754

Contact:

T. Koshy Descriotion: On 03/02/94, South Texas-1 reported that a rapid depressurization could occur in the HVAC ducts in the event of a tornado. A design deficiency limited the closing of a damper. On 10/22/93, River Bend reported in LER 93-020 that a design deficiency could result in loss of ventilation to several buildings after passage of a tornado.

Oriainatina Documents: River Bend Station LER 93-020, dated 10/22/93; South Texas Unit 1 LER 94-003, dated 03/02/94.

Reaulatory Assessment: Even though Tornado dampers were not tested, only one of the 30 dampers failed to operate when tested at South Texas Project. The problem at River Bend was the failure of dampers to open when the fans are running. This problem appears to be generic, since an accident condition is not considered to exist while these dampers are challenged, a reasonable time is available to manually manipulate the dampers and recover the affected systems. Therefore, an IN is adequate response to build the awareness in the industry and for overcoming this vulnerability.

Resolution: IN 96-06, " Design and Testing Deficiencies of Tornado Dampers at Nuclear Power Plants," was issued on 01/25/96 (Accession No. 9601190306).

Completion Date: 01/26/96 GCCA-0081: ASSESSMENT OF CORROSION OF B&W FUEL USED IN 2-YEAR FUEL C1CJ15 TAC No.: M93842

Contact:

E.J. Benner

==

Description:==

TMI-l shut down on 09/08/95 for its scheduled refueling and maintenance outage. The licensee found an unusual build-up of corrosion products on approximately 40 of the 177 fuel assemblies. Some of the fuel assemblies with the corrosion product buildup have clad wall thinning of the outer face pins. The licensee inspections revealed small defects in a total of nine fuel pins in five assemblies. The pin-hole size defects in the pins' metal cladding have occurred only in first cycle fuel with high (4.75%) enrichment.

06/30/96 A.E-62 NUREG-0933

Oriainatina Document: NRC Morning Report I-95-0126, dated 09/27/95.

Reaulatory Assessment: The licensee believes that the corrosion and associated I

defects are due to particulars of this fuel cycle, which was the first two-year l

cycle of the unit. The particulars include highly enriched fuel (to allow a two-year cycle with the lower enrichment once-burned and twice-burned fuel), and the high RCS boron concentration and high power peaking factors associated with highly enriched fuel. Mid-cycle, the licensee instituted a lithium addition program to combat corrosion found in the CRD area (also attributed to the high boron concentration) and believes that this will reduce future corrosion of the fuel assemblies.

The safety significance of this type of fuel cracking is very likely bounded by the conditions that existed during this cycle, since the corrosion was limited to " peaking" fuel assemblies' and future cycles will have lower peaking factors.

This concern may be generic as licensees migrate to two-year fuel cycles. The NRC will continue to follow the licensee corrective action and take action as necessary. TAC No. M93842 was opened for SRXB to evaluate the need for additional generic action.

Resolution: SRXB determined that additional follow-up activity was necessary.

RVIB has been tasked with inspecting the B&W fuel fabrication facility to evaluate manufacturing processes and use of Codes to predict fuel failures during extended cycles.

Completion Date: 02/02/96 ECA-0082: ENVIRONMENTAL EFFECTS ON MAIN STEAM SAFETY VALVE SET POJNJ TAC No.: M94004

Contact:

J.R. Tappert

==

Description:==

On 09/22/95, AN0-2 began testing their main steam safety valves (MSSV) using a Crosby lift assist device. The first valve tested did not lift at a simulated pressure of 6.12% over the nominal set pressure. The licensee stopped testing, reviewed their procedures, and the following day resumed testing. The next valve tested lifted at 4.27% over nominal and the third valve would not lift at 5.9% over nominal. At this point, the licensee stopped testing and developed a detailed action plan. ANO subsequently removed all 10 MSSVs and shipped them to Wyle Laboratories for testing and/or refurbishment.

On 09/30/95, Wyle tested one of the valves ANO was unable to lift in-situ. The valve lifted at 0.97% above nominal which was within the acceptance range. The licensee had personnel onsite at Wyle and immediately began to investigate the discrepancy between the two test results. The difference appears to be caused by the differences in the environments in which the valves were tested. The in-situ test was done with uninsulated valves with an ambient temperature of approximately 95*F. The Wyle test was done with insulated valves in an ambient environment of 140 F. When the valve was retested at Wyle attempting to replicate the ANO environment, the safety valve lifted at 5.65% above nominal, much closer to the in-situ observed conditions. Wyle then went on to test all of the MSSVs O

meet the TS required tolerances of +1%/-3%, with two valves exceeding +6%.

using simulated ANO environmental conditions. Five of the ten valves failed to l

06/30/96 A.E-63 NUREG-0933

Oriainatina Document:

LER 50-368/95-05, dated 11/02/95 (Accession No.

9511080301).

Reaulatory Assessment: Because the licensee did not provide adequate guidance to its vendor, the MSSVs were mis-calibrated. The licensee subsequently determined that using actual as found values for the safety valves, they could show protection for all DBEs. It was not clear how many licensees might also have a discrepancy between their testing and operational environments, so an IN was issued to inform them of AN0's experience.

B3 solution: IN 96-03, " Main Steam Safety Valve Setpoint Variation as a Result of Thermal Effects," was issued on 01/05/96 (Accession No. 9512290299).

Completion Date: 01/15/96 GCCA-0083: INADVERTENT DRAINING OF REACTOR VESSEL AND ISOLATION OF SHUTDOWN COOLING SYSTEM TAC No.: M94044

Contact:

N.K. Hunemuller

==

Description:==

In November 1995, Hatch-2 was in its twelfth refueling outage in the cold shutdown mode; the "A" loop of the RHR system was in the shutdown cooling mode. In accordance with the recent licensee implementation of improved STS, component operation from the remote shutdown panel was being tested for the first time. When maintenance and operations personnel performed activities to determine the cause of deficiencies identified during the testing, approximately 12,000 gallons of water drained out of the reactor vessel in less than 1 minute. The low level of water in the reactor vessel triggered automatic isolation of the shutdown cooling system, terminating the event. Further investigation revealed that an interlock designed to prevent a drain-down had been set improperly, actually causing the event. Although the event was compounded by personnel, procedural, and maintenance errors, NRC inspectors attributed the root cause to inadequate modification, maintenance, and testing control with respect to the remote shutdown panel and related equipment.

Recent implementation of improved STS at various utilities may result in surveillance tests using circuitry that previously went unchallenged. Over time, these circuits may have degraded or were modified and caused unexpected performance. The normal plant configuration may not be the most desirable configuration for these new tests. For example, the normal control switch line-up on the remote shutdown panel may be an appropriate line-up for mitigating a control room fire but may be less appropriate for testing individual components.

The licensee operational experience described in this IN highlights the importance of plant configuration control when implementing new surveillance tests.

Oriainatina Document: Event Notification 29548, dated 11/02/95.

Reaulatory Assessment: The safety significance of this particular event appears to be limited. The shutdown cooling containment isolation functioned as designed and multiple ECCS makeup sources were operable as required by TS. Reactor vessel water level remained above the top of active fuel. However, an IN highlighting both the speed of the draindown and plant configuration control for tests 06/30/96 A.E-64 NUREG-0933

4 i

i i O warranted.

involving operations from the remote shutdown panel was determined to be Resolution: IN 96-15, " Unexpected Plant Performance During Performance of New

}

Surveillance Tests," was issued on 03/08/96 (Accession No. 9603040234).

}

Comoletion Date: 03/13/96 l

j GCCA-0084: RECENT PROBLEMS WITH OVERHEAD CRANES TAC No.: M94045

Contact:

J.R. Tappert j

Descriotion: Problems with overhead cranes had been identified at two different i

sites. At Trojan, a section of the reactor building polar crane bridge rail

]

failed due to the inappropriate flame-cutting of bolt slots during initial construction.

At Prairie

Island, the overhead crane handling system inappropriately automatically stopped on overload while lifting a loaded spent fuel storage cask. The crane stopped due to an inaccurately calibrated overload-sensing system.

Oriainatina Documents: Inspection Reports 50-344/95-06, dated 09/18/95 (Accession No. 9509210219), and 50-282/95-06, dated 06/27/95 (Accession No. 9507070029).

Reaulatory Assessment: Crane failures adversely affect plant operations and could lead to a radiological accident. An IN was promulgated to inform licensees of O

these recent problems.

Resolution: IN 96-26, "Recent Problems with Overhead Cranes," was itsued on 04/30/96 (Accession No. 9604260095).

Completion Date: 04/30/96 GCCA-0085: REMOVING REFUELING FLOOR SHIELDING PLUGS PRIOR TO AND SOON AFTER SHUTDOWN TAC No.: M94088

Contact:

E.Y. Wang Descriotion: Oyster Creek was planning to move the shield plugs before shutdown.

Weighing 10 to 85 tons, the shield plugs are layered directly above the reactor vessel. The primary purpose of shield plugs is to provide protection to plant J

personnel working on the refueling floor. Oyster Creek wanted to save outage time by removing the shield plugs before the reactor was shutdown. Other utilities have similar practices.

Oriainatina Document: Region I SMM Pre-brief in October 1995.

Reaulatory Assessment: A concern was raised regarding moving shield plugs before a plant is shut down, just before a refueling outage begins. There are several sites that have such practice, including Oyster Creek, Limerick, and Millstone-1; Cooper has recently started the same practice. Safety concerns have been raised regarding personnel safety under accident conditions and under LOCA conditions, after the shield plugs are removed. There is also a concern of dropping the heavy 06/30/96 A.E-65 NUREG-0933

1 l

load of shield plugs, a condition which has not been analyzed at some of the plants. At one site, this activity was precluded by the FSAR; however, this activity was done under 10 CFR 50.59. A question exists whether 10 CFR 50.59 I

adequately considered all of the above safety concerns.

l l

Resolution: The resolution of this concern was included in the Bulletin 96-02,

" Movement of Heavy Loads Over Spent Fuel Pool, Over Fuel in the Reactor Core, or l

Over Safety-Related Equipment," issued on 04/11/96 (Accession No. 9604080259).

l TAC M94912 was issued to capture the effort on this bulletin. The subject of the bulletin was to address concerns of moving heavy loads, including dry casks and other heavy loads over reactor vessel and spent fuel pool. Since the concern with shield plugs was addressed in this bulletin, no separate generic communication was deemed necessary.

[gmDietion Date': 05/07/96 GCCA-0086: DAMAGE TO VALVE INTERNALS CAUSED BY THERMALLY-INDUCED PRESSURE LOCKING TAC No.: M94189

Contact:

T.J. Carter

==

Description:==

Observed damage to an internal component of a valve was attributed to thermally-induced pressure locking. A retaining ring had been bent; this was indicative of an internal pressure between 3000 and 7000 psi. Valves adjacent to piping systems subje::ted to large temperature changes could be susceptible to thermally-induced pressure locking, inoperability, and possible damage.

Oriainatino Document: Event Notification 29659, dated 11/30/95.

Reaulatory Assessment: Temperature increase of fluid trapped in valve bonnets can cause very high pressures. Potential pressure locking and thermal binding is being addressed in GL 95-07. A number of improbable factors, such as a heat source sufficient to actually raise the traoped fluid temperature and a valve that is leak-tight enough to contain the resulting pressure increase, are necessary to achieve very high pressures. An IN should be prepared that alerts licensees to the potential for an undetected thermally-induced pressure increase in valves.

Resolution: IN 96-08, " Thermally Induced Pressure Locking of a High Pressure Coolant Injection Gate Valve," was issued on 02/05/96 (Accession No. 9601300092).

Comoletion Date: 02/05/96 GCCA-0087: DAMAGE IN FOREIGN STEAM GENERATOR INTERNALS TAC No.: M94254

Contact:

E.J. Benner Descriptign: In April 1995, during a routine eddy current inspection of the SG tubing at a foreign facility, anomalous support plate signals were observed at the uppermost support plate. The SGs are similar to W Model 51 SGs. The support plates are of the drilled-hole type and are fabricated from carbon steel. Video camera inspections were conducted to investigate the anomalous signals and revealed that a significant portion of the support plate had wasted away. Pieces 06/30/96 A.E-66 NUREG-0933

i

/9 of the affected region of the support plate were found resting on the next lower V

support plate. Subsequent investigation identified chemical cleaning performed in 1992 as the cause of the support plate damage.

Oriainatina Document: Bilateral agreement with a foreign country.

Reaulatory Assessment: The SG tube support plates function to support the tubas against lateral displacement and vibration and to minimize bending moments in the tubes during accidents. Known instances of support plate cracking / damage in the U.S. have generally involved support plates with significant denting. The potential for support plate cracks has tended to not be of significant concern in recent years since the SGs most affected by denting have been replaced and the industry has been successful in controlling denting progression. The foreign experience serves to highlight that there are other mechanisms which may lead to support plate cracking / damage.

Based on the information available to the staff, it is not yet known whether SGs i

in the U.S. are vulnerable to the type of wrapper damage observed at the foreign unit. The staff will continue to monitor information on support plate and wrapper damage as it becomes available from foreign authorities. Issuance of an IN is commensurate with the known safety significance and applicability of the concern.

Factors impacting priority determination were the potentially high safety significance and the lack of direct available evidence indicating applicability to U.S. plants.

Resolution: IN 96-09, " Damage in Foreign Steam Generator Internals," was issued on 02/12/96 (Accession No. 9602060170).

C Completion Date: 02/12/96 GCCA-0088: INTERFACE BETWEEN OPERATORS AND NUCLEAR _ ENGINEERS DURING TESTS AND STARTUP TAC No.: M94370

Contact:

E.M. McKenna

==

Description:==

During individual control rod testing in March 1995, Dresden-3 briefly exceeded the TS limit for maximum fuel design limiting ratio for centerline melt (FDLRC). The nuclear engineer recognized that the limit might be exceeded when a particular (high-worth) rod was withdrawn, but also knew that subsequent insertion of the rod would restore the ratio. The potential exceedance was not communicated to the operators. The licensee investigation showed weaknesses in its reactivity control program such as incorrect focus on pre-conditioning limits and failure to perform predictor model calculations when conditions changed. Further, it was noted that operations personnel heavily i

relied on the nuclear engineers and lacked sufficient knowledge to question thermal limit trends.

Oriainatina Document: Dresden Unit 3 (50-249) LER 95-05-01, dated 11/16/95 (Accession No. 9511210148).

O Reaulatory Assessment: The potential exists that facility personnel other than

\\

NRC-licensed operators may effectively control reactivity manipulations under certain circumstances. An example is a nuclear engineer supervising rod testing, 06/30/96 A.E-67 NUREG-0933

if the licensed operator is positioning rods without having the knowledge or procedures necessary to ensure compliance with applicable TS limits. The safety significance of the specific exceedances at Dresden was low due to the small amount of exceedance and the short duration. However, the Events Assessment Panel authorized issuance of a TAC No. for long-term follow-up of potential generic concerns about the interface between operations and nuclear engineering for reactivity control situations.

Resolution: The Operator Licensing Branch had conducted e review of the adequacy of licensee control of reactivity changes during startup and during rod pattern manipulations as part of another task. As part of this review, an informal survey of reactor engineering and operations practices was performed, and a search of LER data bases. Based on this review, risk insights, and the indication /

protection systems available, it was concluded that generic action by the NRC is not necessary (see memo from S. Richards to A. Chaffee on 02/05/96).

Comoletion Date: 02/05/96 GCCA-0089: VALVE STEM COUPLING OF GIMPEL AUXILIARY FEEDWATER TURBINE TRIP THROTTLE VALVES IAC No. : M94371

Contact:

T.J. Carter

==

Description:==

Reports were received that identified 2 mechanisms for disengagement of linkages used in the operation of turbine governor controls. One involved missing or improperly installed set screws that would prevent unscrewing of a coupling. The other mechanism involved the use of too " thick" a lock nut whose locking mechanism was not engaged.

Oriainatina Documents:

Morning Reports 3-94-0146 and 4-94-0102, Event Notification 29111, and a vendor letter dated 08/30/94 (Accession No.

9409150014).

Reaulatory Assessment: When the connecting coupling between the valve stem and operator. becomes loose, erratic operation of the control valve is observed. This could impact operation of the turbine throttle valve used in the AFW system of PWRs, a safety-related system. The defect also could impact BWRs, both the RCIC and HPCI systems. The priority for resolution'was judged to be moderate even though the probability of failure is believed low based upon the number of observed failures. Based on follow-up information, the importance was less than originally perceived. By this time, the coupling deficiency had been addressed by the vendor notifying their customers.

Resolution: Issuance of an IN was cancelled. The coupling concern was addressed by the vendor.

Comoletion Date: 04/30/96 0

06/30/96 A.E-68 NUREG-0933

i GCCA-0090: IMPROPER EQUIPMENT SETTINGS DUE TO THE I!'.E OF NON-TEMPERATURE 6

COMPENSATED TEST EOUIPMENT TAC No.: M94458

Contact:

E.N. Fields Descriotion: The use of non-temperature compensated test gauges to calibrate and test safety-related equipment was identified at Farley and Surry. Non-temperature compensated gauges were used in environments that required that temperature corrections be applied to gauge readings. However, licensees were not correcting gauge readings.

Oriainatina Document: Memorandum from E. Mershoff to D. Crutchfield, dated 11/20/95.

Reaulatory Assessment: No immediate safety concern was identified, i.e.,

no instances were identified where TS setpoints were exceeded; however, a potential existed that a limit could be exceeded. Systems potentially affected included reactor trip system transmitters, MSSV lift settings, ESFAS transmitters, and pressure instruments used for calorimetric calculations. Therefore, an IN was determined to adequately address this issue.

Resolution:

IN 96-22, "laproper Equipment Settings Due to the Use of Nontemperature-Compensated Test Equipment," was issued on 04/11/96 (Accession No.

9604050336).

Comoletion Date: 05/07/96 O

GCCA-0091: USE OF INDIVIDUAL PLANT EXAMINATIONS (IPEs) FOR REGULATORY DECISION MAKING TAC No.: M94469

Contact:

N.K. Hunemuller

==

Description:==

An IN was proposed to restate the objectives of IPEs, including the IPEEEs, to address the purpose of the staff review of the IPE and IPEEE i

submittals, and to address the potential relationship of the IPE and IPEEE program to other ongoing and future regulatory programs. The proposed IN was to inform licensees that the use of information from an IPE submittal for purposes other than those associated with GL 88-20 would likely require additional staff review.

Oriainatina Document: Memorandum from A. Thadani, "Use of Individual Plant Examinations (IPEs) for Regulatory Decision Making," dated 10/03/95 (Accession No. 9510100018).

Reaulatory Assessment: The majority of licensees have indicated their intention to update and maintain their IPEs (i.e.,

PRAs) and to use these PRAs in regulatory applications beyond GL 88-20. This use of PRA is encouraged in the Commission's Policy Statement on "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities" and the staff is pursuing activities directed toward greater use of risk information in regulatory decision-making. However, O

the staff review of the IPEs has not been performed for this purpose. Therefore, to help ensure that the scope (and limitations) of the staff's review of the IPE submittals is clearly understood, guidance has been provided to the staff to 06/30/96 A.E-69 NUREG-0933

clarify that the staff review was not designed to provide the sole basis for risk-informed regulatory decision-making and that use of a licensee's IPE for regulatory decisions other than associated with GL 88-20 will likely require additional staff review. A letter to the NEI was issued on 04/24/96 to provide similar guidance to licensees.

Resolution: On 04/19/96, a Commission Paper entitled " Clarification of IPE/IPEEE Objectives, Staff Review Purpose, and Potential Future Regulatory Uses of IPE/IPEEE" was issued with a letter to the NEI attached. On 04/24/96, the letter to W. Rasin (NEI) from A. Thadani (NRR) was issued.

Comoletion Date: 03/04/96

_GQCA-0092: OVERWITHORAWAL OF TIP TAC No.: M94470

Contact:

E.Y. Wang Oftic_ tint _ig_q: On 10/31/95, LaSalle-1 experienced difficulty with overwithdrawal of the TIP outside its shield and shield room. This resulted in high radiation levels in portions of the reactor building. The licensee declared an alert based on the radiation presenting a potential over-exposure of plant staff.

Oriainatina Documeq1: Event Notification 29259, dated 10/31/95.

Regulatory Assessment: The TIP system has total of five channels; "lB" was withdrawn beyond t~ne shielded storage location in the reactor building. The LaSalle TS requires at least four operable drive machines and that TIP data for an inoperable measurement location may be replaced by data obtained from a 3-dimensional BWR core simulator code normalized with available operating measurements. Since the other four TIPS were operable, there is no operational problem in with this event. There is an area radiation monitor in the vicinity of TIP drive machines which provide an alarm in the control room. In addition, LaSalle personnel uses electronic dosimeters which also provide alarm when radiation level reaches certain setpoint.

Resolution: IN 96-25, " Traversing In-Core Probe Overwithdrawn at LaSalle County Station, Unit 1," was issued on 04/30/96 (Accession No. 9604250172).

Comoletion Date: 04/30/96 GCCA-0093: SPENT FUEL POOL COOLING TAC No.: M94480

Contact:

D.L. Skeen

==

Description:==

The adequacy of SFP cooling at nuclear power plants was called into question when it was discovered that Millstone-1 routinely performed full-core off-loads during refueling, even though the plant's licensing basis described a partial-core off-load during normal refueling. As part of the subsequent Task Action Plan on Spent Fuel Storage Pools (TAC No. M88094), all NRR project managers were directed to perform a survey of the current licensing basis for the SFP at each plant. TAC No. M94480 was issued to track the time project managers 06/30/96 A.E-70 NUREG-0933

l

{

spent conducting the survey. The results of the survey were submitted to SPLB for l

evaluation.

Oriainatina Document: Memorandum from the ED0 to the Chairman concerning lessons l

learned from Millstone Unit 1, dated 12/28/95 (Accession No. 9603120370).

Reaulatory Assessment: After evaluating the survey results, the staff determined that the existing structures, systems, and components related to storage of irradiated fuel provide adequate protection for public health and safety.

However, the staff review identified strengths and weaknesses and potential areas for safety enhancements for individual plants.

Resolution: The concern was resolved when all project managers submitted their survey results to SPLB (see TAC No. M88094 for resolution of SFP issues).

l Comoletion Date: 05/31/96 GCCA-0094: SOUTH TEXAS STUCK ROD EVENT FOLLOWING REACTOR TRIP TAC No.: M94494

Contact:

S.S. Koenick

==

Description:==

On 12/18/95, with South Texas-1 at 100% power, a pilot wire monitoring relay actuation caused a main transformer lock-out which resulted in l

a turbine trip and reactor trip. In response to the reactor trip, three control rod bottom ligh+.s failed to light and the digital rod position indicated six l

steps out for each rod. Following the transient, one rod drifted to the bottom and the other two were manually inserted. During subsequent rod testing, the three control rods and an additional control rod failed to fully insert.

South Texas has a 14-foot core with ){ Standard XL, Standard XLR, and VANTAGE 5H 17 x 17 fuel assemblies, and the affected control rods were found in twice-burned Standard XLR fuel with burnup greater then 42,880 megawatt days (MWD)/ metric ton uranium (MTU).

Oriainatina Document: Event Notification 29734, dated 12/18/95.

Reaulatory Assessment: On a plant-specific basis, the transient was within the plant design basis and all systems effectively functioned as designed. With respect to the rods stopping six steps from the bottom, the rod worth for the last six steps is minimal and the licensee verified adequate shutdown margin.

On a generic application of control rod problems, the item is significant in that stuck control rods could affect safe shut down margin following design basis transients. For South Texas, the licensee performed a safety evaluation that supports negligible impact to shutdown margin and reload safety evaluation assuming 32 control assemblies do not insert below 12 steps following a reactor trip. Therefore, there appears to be justification for continued operation while the root cause is being pursued.

Resolution: IN 96-12, " Control Rod Insertion Problems," dated 02/15/96 (Accession

, O No. 9602090161), discussed details of both the South Texas trip and the Wolf Creek trip on 01/30/96. Subsequently,Bulletin 96-01 (Accession No. 9603120001) i 06/30/96 A.E-71 NUREG-0933

was issued on 03/08/96 requesting W utilities to conduct specified control rod tests. The rcot cause of the control rod problem is still under investigation.

Completion Date: 02/15/96 GCCA-0095: RADWASTE FACILITY EOUIPMENT DEGRADATION AT MILLSTONE UNIT 1 TAC No.: M94521 Centact: E.Y. Wang Descriotion: During an NRC routine inspection, it was identified that a portion of the radwaste processing facility was apparently not maintained; the waste storage tanks had indications of leaks, there was indication of a few feet of flooding in the waste storage room; corrosion of piping and tanks in the facility was evident; and waste build-up and equipment damage were also observed.

Oriainatina Document: Inspection Report 50-245/95-35, dated 09/11/95 (Accession No. 9509180214).

fLqaulatory Assessment: The radwaste room condition has degraded significantly because of the waste build-up and equipment damage. The radwaste room is designet to have limited access. The radwaste storage tank has degraded to a condition that it no longer could store the waste. The whole room became a storage place for the radwaste. The r perational safety significance concern is minimal. There i

is no specific NRC refulation regarding the radwaste room conditions. Yet, NRC issued IE Circular No. 80-18 on 08/22/80 (Accession No. 8006190038), "10 CFR 50.59 Evaluations For Changes To Radioactive Waste Treatment Systems," which requires licensee to perform a safety eva~luation in accordance with 10 CFR 50.59 for the following circumstances: (1) components described in the SAR are removed; (2) component functions are altered; (3) substitute components are utilized; or (4) changes remain following completion of maintenance activity.

Resolution: IN 96-14, " Degradation of Radwaste facility Equipment at Millstone Nuclear Power Station, Unit 1,"

was issued on 03/01/96 (Accession No.

9602260117)

Comoletion Date: 03/04/96 GCCA-0096: WOLF CREEK REACTOR TRIP WITH ONE TRAIN ESSENTIAL SERVICE WATER SYSTE3 INOPERABLE

, TAC No.: M94594

Contact:

J.R. Tappert

==

Description:==

On 01/30/96, operators at Wolf Creek received alarms indicating that the circulating water system traveling screens were becoming blocked. A visual inspection showed that the traveling screens for Bays 1 and 3 were frozen and that water levels in these bays ware approximately 8 ft below normal. The essential service water system was started with the intent to separate this system from the service water system. At approximately 3:30 a.m.,

operators received a service water pressure alarm and an electric fire pump started on low service water pressure.

The shift supervisor then directed a manual reactor / turbine trip. Circulating water system bays were subsequently determined to be at 12 feet below normal. The level loss was caused by water from the spray wash system freezing and blocking the traveling screens.

06/30/96 A.E-72 NUREG-0933

V]

(

The Train A essential service water system pump was tripped and declared inoperable at 7:47 a.m. because of low discharge pressure and high strainer differential pressure. At about 5:45 p.m.,

the operators declared Train A operable on the basis of an engineering evaluation and placed it in service.

However, the pump was again stopped 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> later at approximately 7:30 p.m.

when the pump exhibited further oscillations in flow and pressure. At approximately 8:00 p.m., operators noted that essential service water system Train B suction bay level was 15 ft below normal and decreasing slowly.

Operators placed additional heat loads on Train B and the suction bay levels subsequently recovered.

At about 10:00 a.m. on 01/31/96, divers inspected the suction bay of Train A and noted complete blockage of the trash racks by frazil ice. Train B was not inspected because the pump was running. The ice blockage was cleared by 4:00 p.m.

by sparging the trash racks with air. The essential service water system was designed to have warming flow injected in front of trash racks to increase bulk water temperature and prevent the formation of frazil ice. Due to calculational errors by the architect-engineer and the as-built system configuration, the essential service water system warming flow was insufficient to prevent frazil ice from forming at the Train A trash racks.

Oriainatina Documents: Event Notification 29904 and 29905 dated 01/30/96, LER 50-482/96-01 (Accession No. 9603120274), and LER 50-482/96-02 (Accession No.

960310619).

Reaulatory Assessment: Facility vulnerability to icing events is a function of n

)

plant design. Frazil and other ice formation is dependent on specific (V

environmental conditions and represent a potential common-mode failure that can cause the loss or degradation of iaultiple cooling water systems, including the potential loss of the UHS. Loss of the UHS is potentially significant and it was not clear what facilities would be vulnerable to this failure mode. Therefore, an IN was issued to alert licensees to recent ice-related events.

Resolution: IN 96-36, " Degradation of Cooling Water Systems Due to Icing," was issued on 06/12/96 (Accession No. 9606070097).

Comoletion Date: 06/12/96 GCCA-0097: STUCK CONTROL ROD PROBLEMS TAC No.: M94608

Contact:

J.R. Tappert Descriotion: PWR control rods fail to completely insert upon a scram signal. On l

12/18/95, with South Texas-1 at 100% percent power, an electrical transient led to a reactor trip. While verifying that control rods had inserted fully after the l

trip, operators noted that the rod bottom lights of 3 control rod assemblies were l

not lit; the digital rod position indication for each rod indicated 6 steps l

withdrawn. A step is equivalent to 1.59 cm (5/8 inch] and the top of the dashpot begins at 38 steps. During subsequent testing of all control rods in the affected banks, the rod position indication for the same 3 locations as well as a new

(]

location indicated 6 steps withdrawn. As compared to prior rod drop testing, no V

significant differences in rod drop times were noted before reaching the upper dashpot area for any of the control rods. All 4 control rods were located in fuel 06/30/96 A.E-73 NUREG-0933

I assemblies that were in their third cycle 'with burnup greater than 42,880 megawatt days per metric ton uranium (MWD /MTU).

On 01/30/96, after a manual scram from 80% power, 5 control rod assemblies at Wolf Creek failed to insert fully. Two rods remained at 6 steps withdrawn, 2 at 12 steps, and 1 at 18 steps. At Wolf Creek, a step is equivalent to 1.59 cm [5/8 inch) and the top of the dashpot begins at approximately 30 steps. Three of the affected rods drifted to fully inserted within 20 minutes, I within 60 minutes, and the last one within 78 minutes. The results also indicate that there was some slowing down of affected rods before reaching the dashpot. During subsequent cold drop tests, the same 5 rods plus an additional 3 rods failed to fully insert.

All of the affected rods were in 17x17 VANTAGE 5H fuel assemblies with burnup greater than 47,600 MWD /MTV.

On 02/2)/96, during the insert shuffle in preparation for loading North Anna-1 Cycle 11, 2 new control rods assemblies could not be removed with normal operation of the handling tool from the fuel assemblies in the spent fuel pool in which they were temporarily stored. The control rod assemblies were removed using the rod assembly handling tool in conjunction with the bridge crane hoist.

The 2 affected fuel assemblies were VANTAGE 5H assemblies which had achieved 47,782 MWD /MTV and 49,613 MWD /MTV burnup during 2 cycles of irradiation.

Oriainatina Documents: PNO-IV-95-059 and Event Notification 29904, dated 01/30/96.

Reaulatory Assessment: The events discussed earlier, as well as several similar events at foreign reactors, raise concerns about the operability of control rods in high burnup fuel assemblies. While most of the testing to date has demonstrated that the control rods have reached the dashpot region of the guide tube and that adequate shutdown margin has been maintained, there have been indications of degraded rod drop times and a stuck rod well above the dashpot region. Thus, there is concern that these events may be precursors to more significant control rod binding problems in which required shutdown margins and rod drop times may be violated. Due to the fact that the control rod binding mechanism was not fully understood a Bulletin was issued to alert licensees to these events and request that licensees with H-designed plants assess control rod operability and verify the control rod drop times, rod recoil, and drag forces at the next scheduled shutdown (E0C, maintenance, etc.) for all rodded fuel assemblies.

Resolution: Bulletin 96-01, " Control Rod Insertion Problems," was issued on 03/08/96 (Accession No. 9603120001). Also, IN 96-12, " Control Rod Insertion Problems," was issued on 02/15/96 (Accession No. 9602090161).

Completion Date: 03/13/96 0

06/30/96 A.E-74 NUREG-0933

l.

O GCCA-0098:

FAILURE OF TONE ALERT RADIO TO ACTIVATE WHEN RECEIVING A SHORTENED ACTIVATION SIfdlAL TAC No.: M94768

Contact:

J.B. Birmingham

==

Description:==

During an Emergency Rtsponse Test at Callaway, the length of the tone alert signal time was found to be insufficient to activate some portions of the Tone Alert Network.

Oriainatina Document: IN Authorization Request Form presented to the Events Assessment Panel on 02/13/96.

Reaulatory Assessment: This concern has a moderate degree of safety significance ana generic applicability in that tone alert radio signals are typically part of an overlapping system designed to alert the general populace in the event of needed emergency action. Although the FCC authorized the time length of the signal to be reduced, many stations have not made any changes. Additionally, the failure of the tone to activate tone alert radios could be detected during emergency response tests. However, the potential failure of the signal to activate tone alert radios reduces the overall performance of the alert responsa system.

Resolution: IN 96-19, " Failure of Tone Alert Radio to Activate when Receiving a Shortened Activation Signal," was issued on 04/02/96 (Accession No. 9603270127).

Comoletion Date: 04/02/96 O

ECA-0099: SLOW FIVE PERCENT SCRAM INSERTION TIMES CAUSED BY VITON DIAPHRAGMS IN SCRAM SOLEN 0ID PILOT VALVES TAC No.: M94778

Contact:

D.L. Skeen

==

Description:==

Degradation of the control rod 5% scram insertion times has been noted at some BWR plants. GE and the manufacturer, Automatic Switch Company (ASCO), have determined that the cause of the slow times is adherence of the fluoroelastomer (Viton) diaphragm to the brass valve seat of the scram solenoid pilot valve (SSPV).

Oriainatina Document: Event Notification 29879 from Brunswick Unit 1, dated 01/23/96.

Reaulatory Assessment: Scram time limits are imposed by TS to ensure that the control rods will be inserted into the reactor core fast enough to prevent i

exceeding the minimum critical power ratio (MCPR) and, thus, prevent damage to the fuel. At Brunswick, the TS limit for average core-wide insertion to notch 46 (or 5% into the core) for all control rods is 0.358 seconds. Scram data taken by the licensee following the manual scram on 01/23/96 showed that the TS limits were exceeded (actual core-wide average was about 0.4 seconds).

The event scenario of concern is a reactor trip at end of core life without turbine bypass valves opening. At the end of core life all control rods are fully

(

withdrawn from the core (Position 48) and the flux pattern is shifted to the top N

of the core. Thus, a delay in scram time could potentially cause some fuel to be 06/30/96 A.E-75 NUREG-0933

damaged. However, GE performed a safety analysis and determined that exceeding the 5% insertion time would not result in core damage as long as the other TS-required scram insertion times (20%, 50%, and 90%) were met. None of these other insertion times have been exceeded as a result of the diaphragm sticking to the seat.

Resoluti,n: IN 96-07, " Slow 5% Scram Insertion Times Caused by Viton Diaphragms in Scran Solenoid Pilot Valves" was issued on 01/26/96 (Accession No.

9601260139). After meeting with the NRC on 01/26/96, the BWR Owners Group activated their Regulatory Response Group to resolve the issue. GE and the BWROG workod with ASCO to develop an improved formulation of Buna-N rubber as an interim measure to replace the Viton diaphragm. The Buna-N diaphragm became available in June 1996. GE is currently developing a composite diaphragm that will include the original Viton as the outer portion because of its superior elastic quality with a center portion made of a much harder Viton to prevent the diaphragm from sticking to the valve seat. GE hopes to have the composite diaphragm available by late August 1996.

Comoletion Date: 01/26/96 GCCA-0100: POTENTIAL CLOGGING 0F HPSI THROTTLE VALVES DURING CONTAINMENT SUMP fECIRCULATION PHASE TAC No.: M94808

Contact:

E.J. Benner

==

Description:==

Northeast Utilities determined that eight manual throttle valves in the high-pressure safety injection lines were susceptible to clogging during the recirculation phase of a LOCA at Millstone-2. The licensee based this determination on the fact that the cpenings in the containment sump screens are 0.187" and the minimum dimension within the valve flow path is 0.125".

Oriainatino Document: Event Notification 29999, dated 02/20/96.

Reaulatory Assessment: The manual throttle valves are inaccessible during an accident. The safety significance is exacerbated by the fact that the normal lineup for recirculation at this unit has the low-pressure safety injection pump feeding the HPSI system, with all recirculation flow passing through the HPSI system. The licensee has adopted this arrangement because of structural and vibrational concerns with the LPSI system.

This concern is generic. The Millstone concern was discovered because of licensee review of a similar concern at Diablo Canyon. The concern at Diablo Canyon was dispositioned as not safety significant because the screen was sufficient to prevent a clogging problem. The appropriate NRC action is to expedite issuance of an information notice, in addition to continued evaluation of the issue. No additional generic action is necessary because of several mitigating factors including: (1) particles passing through the sump strainers may be pulverized by the high-pressure safety injection pumps; (2) differential pressure across the valve may force debris through the valve; (3) other plants may have sump strainer openings smaller than the valve opening; (4) debris may settle in the sump at the flow rates involved with high pressure recirculation; and (5) post-accident recirculation lineup (i.e., whether all recirculation flow must pass through the HPSI system).

06/30/96 A.E-76 NUREG-0933

O, Resolution: IN 96-27, " Potential Clogging of High Pressure Safety Injection Throttle Valves During Recirculation," was issued on 05/01/96 (Accession No.

9604260077).

Comoletion Date: 05/01/96 GCCA-0101: STEAN GENERATOR TUBE INSPECTION RESULTS TAC No. M94862

Contact:

E.J. Benner 1

i Descriotion: SG tube examinations have been performed at a number of plants during the last year. As a result of these examinations, degradation has been observed at a number of locations including dented locations, the expansion transition region, free span region, and in the tubesheet crevice.

In addition to identifying a variety of degradation mechanisms, a number of technical concerns have arisen as a result of these examinations with respect to 4

)

classifying inspection results, periodicity of examinations, and expansion of the j

initial inspection scope.

Originatina Documents: Several Event Notifications including: 28446 (Braidwood),

i dated 03/02/95; 28482 (Maine Yankee), dated 03/04/95; 28646 (Kewaunee), dated 04/07/95; 29494 (Diablo Canyon), dated 10/22/95; and 29571 (Byron), dated 11/07/95.

)

h Reaulatory Assessment: In general, the degradation modes observed have been consistent with past experience; however, the results indicate the importance of i

performing comprehensive SG tube examinations using appropriate inspection techniques. SGTRs can provide a direct release path for contaminated primary coolant to the environment via the secondary side safety and relief valves.

Accumulation of water in the SG secondary side can also lead to an overfill condition which can severely aggravate the radiological consequences and increase the likelihood of subsequent failures. An IN is an appropriate generic action to 1

inform licensees of the particulars of recent examination methodologies and failure mechanisms.

4 Resolution: IN 96-38, "Results of Steam Generator Tube Examinations," was issued 4

on 06/21/96 (Accession No. 9606180338).

i Comoletion Date: 06/21/96 GCCA-0102: REACTOR OPERATION BELIEVED TO BE INCONSISTENT WITH THAT DESCRIBED IN THE FSAR TAC No.: M94911

Contact:

T.J. Carter

==

Description:==

Licensees may not be maintaining and operating their facilities in compliance with their licenses and their bases. A staff follow-up to a 10 CFR 2.206 Petition involving Millstone-1 discussed this situation. A self-assessment O

was performed by Northeast Utilities Service Companies (Accession M. 9603150021) in response to a staff order issued on 12/13/95 (Accession No. 9512150278).

06/30/96 A.E-77 NUREG-0933

Oriainitina Documents: A petition from Galatis and Hadley, dated 08/21/95 (Accession No. 9509080209), and its Supplement, dated 08/28/95 (Accession No.

9509110306).

Reaulatory Assessment: The staff has t concern about Northeast Utilities' performance regarding operation, contro? ling facility changes, and maintaining an accurate updated FSAR for Millstone-l. The licensee's self-assessment stated that a potential exists that similar configuration management conditions may exist at several of their other units. To alert other licensees to what may be a generic deficiency, an IN should be issued that transmits both the Millstone licensee's self evaluation and the staff 10 CFR 50.54(f) letter that expresses the concern.

Resolution: IN 96-17, " Reactor Operation Inconsistent With The Updated Final Safety Analysis Report," was issued on 03/18/96 (Accession No. 9603150213).

Completion Date: 03/18/96 GCCA-0103: MOVEMENT OF DRY STORAGE CASKS OVER SPENT FUEL. FUEL IN THE REACTOR CORE. OR SAFETY-RELATED EQUIPMENT TAC No.: M94912

Contact:

E.Y. Wang i

Descriotion: A licensee was planning to move an unanalyzed load along a path where a load drop would have significant safety consequences.

Oriainatina Document: See TAC No M94088.

Reaulatory Assessment: The licensee's 10 CFR 50.59 evaluation was based on a misunderstanding of the purpose of GL 85-11 " Completion of Phase II of ' Control of Heavy Loads at Nuclear Power Plants' NUREG-0612." The staff determined that this was an unreviewed safety question because: (1) the casks were heavier than those previously considered in the FSAR; and (2) a load drop could result in consequences that are greater than previously evaluated in the FSAR and, therefore, the margin of rafety could be reduced.

Resolution: Bulletin 96-02, " Movement of Heavy Loads Over Spent Fuel Pool, Over Fuel in the Reactor Core, or Over Safety-related Equipment," was issued on 04/11/96 (Accession No. 9604080259). The bulletin was to address concerns of moving heavy loads, including dry casks, over the reactor vessel and the spent fuel pool. Since the concern with shield plugs (TAC No. M94088) was addressed in this bulletin, no separate generic communication was deemed necessary.

Completion Date: 05/07/96 CCCA-0104: INACCURACY OF DIAGNOSTIC EQUIPMENT FOR MOTOR-OPERATED ljuTTERFLY VALVES TAC No.: M95281

Contact:

T.A. Greene Descriotion: ITI NOVATS Inc. developed the Butterfly Analysis and Review Test (BART) System as a method for determining the torque output of Limitorque HBC gear boxes equipped with Limitorque motor actuators on butterfly valves.

06/30/96 A.E-78 NUREG-0933

1

/G Observations and questions concerning the performance of the BART System under Q

field conditions led ITI M0 VATS to perform testing to determine more precisely the inaccuracy of the system. This TAC No. was for the issuance of an IN to alert licensees to the increased inaccuracy of the BART diagnostic equipment for measuring torque when operating butterfly valves.

Oriainatina Document: Morning Report 4-96-0042, dated 04/24/96.

Reaulatory Assessment: Testing has raised questions concerning the inaccuracy assumed for ITI M0 VATS BART diagnostic equipment for butterfly valves. The inaccuracy could be as high as 14% compared to the previous 2% error assumption.

The increased inaccuracy could adversely affect safety-related butterfly valves set up with the diagnostic equipment.

Resolution: IN 96-30, " Inaccuracy of Diagnostic Equipment for Motor-0perated Butterfly Valves," was issued on 05/21/96 (Accession No. 9605160311).

Comoletion Date: 05/21/96 GCCA-0105: CROSS-TIED SAFETY INJECTION ACCUMULATORS TAC No.: M95282

Contact:

J.R. Tappert Descriotion: Many licensees may have operated with cross-tied SI accumulators in an unanalyzed condition. On 03/08/96, the licensee for Indian Point-3 (IP-3)

O periodically cross-tied SI accumulators for short periods of time.

reported that they had operated outside of their design basis because they had IP-3 TS actually require periodic cross-connecting under certain conditions. An evaluation by the licensee's engineering staff (confirmed by W) shows that the plant may not be protected during some LOCAs with a cross-tied configuration.

This is because nitrogen pressure is postulated to bleed off through the faulted loop to the containment. After the IP-3 report, several other licensees including IP-2, Turkey Point-3 & 4, Byron-1 & 2, Braidwood-1 & 2, Zion-1 & 2, and Vogtle-1

& 2 reported that their plant procedures allow cross-connection of SI accumulators (in some cases, all of the accumulators) in order to equalize pressure. No other licensee reported a requirement to perform this operation.

Oriainatina Documents: Event Notification 30364, dated 04/25/96, initiated action, but the first notification was from IP-3 Event Notification 30087, dated 03/08/96.

Reaulatory Assessment: IP-3 performed a probabilistic evaluation to determine if the configuration exceeded the EPRI screening criteria of 10E-6 core damage probability. The licensee concluded that, for the bounding time estimate of 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> / year for cross-connecting two SI accumulators, the EPRI screening threshold was not met. IP-3 has taken actions to change their TS and other licensees have taken administrative action to prohibit cross-connecting the accumulators.

Therefore, this concern appears to be generic, but the safety significance is limited and the licensees who have been made aware of the problem have taken corrective action.

Resolution: IN 96-31, " Cross-Tied Injection Accumulators," was issued on 05/22/96 (Accession No. 9605170288) to alert licensees to the potential problem.

06/30/96 A.E-79 NUREG-0933

i Comoletion Date: 05/22/96 GCCA-0106: HYDROGEN GAS IGNITION DURING WELDING OF A VSC-24 MULTI-ASSEMBLY SEALED BASKET TAC No.: M95483

Contact:

T.A. Greene

==

Description:==

A hydrogen generation and ignition event occurred at Point Beach during the welding of the shield lid on a spent fuel storage cask. The hydrogen was generated by a chemical reaction between the cask materials and the borated spent fuel water. A review of previous cask loadings at Point Beach and other plants indicates that this situation has occurred before with casks of the same design. This TAC was initiated for the issuan'ce of an IN to alert licensees to the Point Beach event.

Oriainatino Document: Event Notification 30552.

Reaulatory Assessment: The Point Beach event raised concerns about the potential for chemical or other reactions between cask materials,

contents, and environments in spent fuel storage cask designs. The basic concern is that these reactions may create hazardous operating conditions and degrade cask safety components to the extent that the cask's ability to store fuel safety will be compromised.

Resolution: IN 96-34, " Hydrogen Gas Ignition During Closure Welding of a VSC-24 Multi-Assembly Sealed Basket," was issued on 05/31/96 (Accession No. 9605310132).

Comoletion Date: 05/31/96 0

06/30/96 A.E-80 NUREG-0933

R For<u 335 U S. NUCLE AR REGULATORY COMMi$SION 1.

P TN ER E3E BIBLIOGRAPHIC DATA SHEET

~"

(see instructions on the reverse)

2. TITLE ANO SUBTITLE A Prioritization of Generic Safety Issues 3.

DATE REPORT PUBLISHED MONTH YEAR December 1996

4. FIN OR GRANT NUMBER

6. AUTHOR (S)

6. TYPE OF REPORT R. Emrit, et al.

1. PE RIOD COV E R E D I nsouso.e osteep 01/01/96 to 06/30/96

8. P F GANIZATlON - N AME ANO ADOR ESS tilNnc, provide Okukn, Office er nennon, U.S Necka neewteray commiuton, and meetsne addren,11senerector, ameesw Division of Engineering Technology Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

s. Se,OggRG ANIZATION - N AM E AND ADOR ESS tst unc, rrae " sew es e6o.e ;steentraceer, prornse anc o*isJon, ottice er neeson, u.S Nucasa neeva ory ce%

Division of Engineering Technology Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

10. SUPPLEMENTARY NOTES

11. ABSTR ACT (200 worsk w arms The report presents the safety priority ranking for generic safety issues reltted to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The safety priority rankings are HIGH, MEDIUM, LOW, and DROP, and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolution of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative.

12. K E Y WOR DS/DESCRtPTORS tuse mens erpe==s sner war sesdat rewmam Jn aererine the swam s t 3. Ay AlLAB4U1 y 5TATIMENT Unlimited generic safety issues a u cua' " c'^ * * 'ca =

tTAts Papel Unclassified

[

t The neoorr)

(

Unclassified

15. NUMBER OF PAGES

16. PRICE NAC FORM 336 (249)

t I

9 l

l l

l I

O Printed on recycled paper O

Federal Recycling Program