ML20211B888
| ML20211B888 | |
| Person / Time | |
|---|---|
| Issue date: | 09/30/1986 |
| From: | Emrit R, Milstead W, Pittman J, Riggs R, Vandermolen H Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0933, NUREG-0933-S05, NUREG-933, NUREG-933-S5, NUDOCS 8610210214 | |
| Download: ML20211B888 (164) | |
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o UNITED STATES 8' o NUCLEAR REGULATORY COMMISSION
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WASHING ton. D. C. 20555 U'(%,...../
i September 1986 i SUPPLEMENT 5 TO NUREG-0933 "A PRIORITIZATION OF GENERIC SAFETY ISSUES" i
REVISION INSERTION INSTRUCTIONS i
Remove Insert Introduction pp. 21 to 52, Rev. 4 pp. 21 to 52, Rev. 5 pp. 53 to 55, Rev. 2 pp. 53 to 55, Rev. 3 pp. 57 to 60 pp. 57 to 60, Rev. 1 1
Section 1 pp. 1.III.D.3-1 to 7 pp. 1.111.0.3-1 to 8, Rev 1 Section 2 p. 2.C.4-1 p. 2.C.4-1, Rev. 1
- p. 2.C.5-1 p. 2.C.5-1, Rev. 1
- p. 2.C.6-1 p. 2.C.6-1, Rev. 1 Section 3 pp. 3.3-1 to 2 pp. 3.3-1 to 2. Rev. 1
- p. 3.21-1 p. 3.21-1, Rev. 1 i
' pp. 3.36-1 to 5, Rev. 1 pp. 3.36-1 to 5, Rev. 2
- p. 3.74-1 pp. 3.74-1 to 4, Rev. 1 pp. 3.99-1 to 5 pp. 3.99-1 to 5, Rev. 1 pp. 3.114-1 to 3 pp. 3.122-1 to 24 pp. 3.124-1 to 2 pp. 3.125-1 to 6 References pp. R-1 to 50, Rev. 2 pp. R-1 to 59, Rev. 3 i
i i
s 8610210214 860930 PDR NUREG 0933 R PDR
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j TABLE II i
LISTING OF ALL TMI ACTION PLAN ITEMS. TASK ACTION PLAN ITEMS, NEW GENERIC ISSUES. AND HUMAN FACTORS ISSUES This table contains the priority designations for all issues listed in this report. For those issues found to be covered in other issues, the appropriate notations have been made in the Safety Priority Ranking column, e.g., I.A.2.2 in the Safety Priority Ranking column means that Item I.A.2.6(3) is covered in Item I.A.2.2. For resolved issues that have resulted in new requirements for operating plants, the appropriate multi plant licensing action number is listed. 'The licensing action numbering system bears no relationship to the numbering systens used for identifying the prioritized issues. An explanation of the classification and status of t'e issues is provided in the 1 legend below.
legend 4
NOTES: 1 - Possible Resolution Identified for Evaluation
, 2 - Resolution Available (Documented in NUREG, NRC Memorandum, SER, or
.! equivalent)
) h3 3 - Resolution Resulted in eitheri (a) The Establishment of New Regulatory 1 Requirements (By Rule, SRP Change, j or equivalent) or (b) No New Requirements
, 4 - Issue to be Prioritized in the Future j $ - Issue that is not a Generic Safety Issue but should be Assigned j Resources for Completion HIGH - High Safety Priority MEDIUM - Medium Safety Priority LOW - Low Safety Priority i' DROP - Issue Dropped as a Generic Issue
, E - Environmental Issue
! HFPP - Human Factors Program Plan j I - TMI Action Plan Item With Implementation of Resolution Mandated by j NUREG-073788 LI - Licensing Issue MPA - Multi-Plant Action (See Status in NUREG-0748)5
- j{ NA - Not Applicable l po RI - Regulatory Impact Issue y,
- j USI - Unresolved Safety Issue ag I C3 ""
UD n 23 c)
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1 LM 1
i
O CD TAbli 11 (Continued) o N
00 Actibn Lead Lead Office / Safety latest O Plan Item / SPEB Division / Priority latest I ssuanc e MPA issur hn Title Engineer B ranc h Ranking Revision Date No.
.lMI ACTION PLAN ITEMS IA OPERATING PERSONNEL iA L Operating Personnel and Staffing
- 1. A. L .1 Shift Technical Advisor -
NRR/DHFS/LQB I F-01
- 1. A . I . 2 Shift Supervisor Administrative Duties -
NRR/DHFS/LQB I
- 1. A . I . 3 Shift Manning -
NRR/DHFS/LQB I F-02
- 1. A . I . 4 Long-Term Upgrading Colmar RES/DF0/HFBR NOTE 3(a) 1 6/30/84 1.A.2 Training and Qualifications of Operating Personnel I . A. 2.1 Immediate Upgrading of Operator and Senior Operator - - -
Training and Qualifications
- 1. A. 2. l( l) Qualifications - Experience -
NRR/DHFS/LQB I 3 12/31/85 F-03 1.A.2.l(2) Training -
NRR/DHFS/tQB I 3 12/31/85 F-03
- 1. A. 2. l( 3) Facility Certification of Competence and Fitness of -
NRR/DHFS/LQB 1 3 12/31/85 F-03 N Applicants for Operator and Senior Operator Licenses N 1. A. 2. 2 Training and Qualifications of Operations Personnel Colmar NRR/DHFS/LQB NOTE 3(b) 3 12/31/85 NA 1 A. 2. 3 Administration of Training Programs -
NRR/DHFS/LQB 1 3 12/31/85
- 1. A. ?. 4 NRR Participation in Inspector Training Colmar NRR/DHF S/LQB L1 (NOTE 5) 3 12/31/85 NA I . A. 2. 5 Plant Drills Colmar NRR/DHFS/LQB NOTE 3(b) 3 12/31/85 NA
- 1. A . 2. 6 Long-Term upgrading of Training and Qualifications - - -
- 1. A. 2 6( 1) Revise Regulatory Guide 1.8 Colmar NRR/DHFS/LQB HF01.1.2 3 12/31/85 NA
- 1. A 2. 6(2) Staff Review of NRR 80-117 Colmar NRR/DHFS/LQB NOTE 3(b) 3 12/31/85 NA 1.A.2.6(3) Revise 10 CF R 55 Colmar NRR/DHFS/LQB 1. A. 2. 2 3 12/31/85 NA 1.A.2.6(4) Operator Workshops Colmar NRR/DHFS/LQB NOTE 3(b) 3 12/31/85 NA
- 1. A. 2. 6( 5 ) Develop Inspection Proceaures for Training Program Colmar NRR/DHFS/LQB NOTE 3(b) 3 12/31/85 NA 1.A.2.6(6) Nuclear Power Fundamentals Colmar NRR/DHFS/LQB DROP 3 12/31/85 NA 1.A 2.7 Accreditation of Training Institutions Colmar NRR/DHFS/LQB NOTE 3(b) 3 12/31/85 NA 1.A.3 Licensing and Requalification of Operating Personnej
- 1. A. 3.1 Revise Scope of Criteria for Licensing Examinations Emrit NRR/DHFS/LQB 1 4 12/31/85
- 1. A. 3. 2 Operator Licensing Program Changes Emrit NRR/DHF S/0LB NOTE 3(b) 4 12/31/85 NA
= 1. A. 3. 3 Requirements f or Operator Fitness Colmar RES/DRA0/HFSB HIGH 4 12/31/85 C: 1.A.3.4 Licensing of Additional Operations Personnel Thatcher NRR/DMFS/LQB NOTE 3(b) 4 12/31/85 NA
$ 1. A 3. 5 E stablish Statement of Understanding with INPO and DOE Thatcher NRR/DHF S/HF E B LI (N01E 3) 4 12/31/85 NA [
C) up 1 A.4 Simulator Use and Development 7h 1.A.4.1 Initial Simulator Improvement - - -
U l . A. 4.1( 1) Short-lerm Study of Training Simulators Thatcher NRR/DHF S/0LB NOTE 3(b) ? 12/31/85 NA $
!.A.4.l(2) Interim Changes in Training Simulators Thatcher NRR/DHFS/0LB NOTE 3(a) 2 12/31/85 Long-Term Training Simulator Upgrade 1.A.4.2 - - -
- 1. A 4. 2(1) Research on Training Simulators Colmar NRR/DHFS/OtB HIGH 2 12/31/85 O O O
a . -. - ~ ~.. --_~ - - . - - , - , - _ . -- - - . - . - ~ . - _ - . - - . ~ . . . . - . _ _ - - . . _ ~ , ~
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o TABtf II (Continued)
) N l OD Action Lead Lead Of fice/ Safety
- tatest Plan Item / SPE8 Division / Priority latest Issuance MPA l!t Issue No. Title Engineer Branch Ranking Revision Date No.
5
, I.A.4.2(2) Upgrade Training Simulator Standards Colmar RES/DF0/HFBR NOTE 3(a) 2 12/31/85
+
I.A.4.2(3) Regulatory Guide on Training Simulators Colmar RES/DF0/HF BR NOTE 3(a) 2 12/31/85 1.A.4.2(4) Review Simulators for Conformance to Criteria Colmar NRR/DHF5/0LB HF01.3.3 2 12/31/85 NA I.A.4.3 Feasibility Study of Procurement of NRC Training Colmar RES/DAE/RSRB LI (NOTE 3) 2 12/31/85 NA Simulator i I.A.4.4 Feasibility Study of NRC Engineering Computer Colmar RES/DAE/RSRB LI (NOTE 5) 2 12/31/85 NA Q
SUPPORT PERSONNEL
)
) I. B. I Management for Operations J
I . B. l.1 Organization and Management Long-Tern Improvements - - -
I.8.1.I(1) Prepare Draft Criteria Colmar NRR/DHF5/LQB HF01.6.1, 2 12/31/85 NA HF01.6.3
, I.B.I.l(2) Prepare Commission Paper Colmar NRR/DHF5/LQB HF 01. 6.1, 2 12/31/85 NA 4
] HF01.6.3 I.8.1.l(3) Issue Requirements for the Upgrading of Management and Colmar NRR/DHF5/LQB HF01.6.1, 2 12/31/85 NA l Technical Resources HF01.6.3 l g I.B.I.l(4) Review Responses to Letermine Acceptability Colmar NRR/DHF5/LQB HF01.6.1, 2 12/31/85 NA oo HF01.6.3
- I.B.I.l(5) Review Implementation of the Upgrading Activities Colmar OIE/DQASIP/ORP8 NOTE 3(b) 2 12/31/85 NA j I.8 1.1(6) Prepare Revisions to Regulatory Guides 1.33 and 1.8 Colmar NRR/DHF5/LQB HF01.1.2, 2 12/31/85 NA 4
4 HF01.3.4 '
75 l 1.8.1.l(7) Issue Regulatory Guides 1.33 and 1.8 Colmar NRR/DHF5/LQB HF01.1.2, 2 12/31/85 NA HF01.3.4, 75 I . 8.1. 2 Evaluation of Organization and Management Improvements - - -
of Near-Tern Operating License Applicants I.8.1.2(1) Prepare Draft Criteria -
NRR/DHF5/LQB I I.8.1.2(2) Review Near-Tern Operating License Facilities -
NPR/DHF5/LQB I j I.B.I.2(3) Include Findings in the SER for Each Near-Term -
NRR/DL/ ORA 8 I
- Operating License Facility I . 8.1. 3 Loss of Safety Function - - -
1.8.1.3(1) Require Licensees to Place Plant in Safest shutdown Sege RES LI (NOTE 3) 2 12/31/85 NA Cooling Following a Loss of Safety Function Due to Personnel Error
- o I.8.1.3(2) Use Existing Enforcement Options to Accomplish Safest shutdown Cooling-Sege RES LI (NOTE 3) 2 12/31/85 NA j Jo Qs I.B.I.3(3) Use Non-Fiscal Approaches to Accomplish Safest Shutdown Cooling Sege RES LI (NOTE 3) 2 12/31/85 NA $
O
- W m .
w I.B.2 Inspection of Operatino Reactors =**
W I . 8. 2.1 Revise OIE Inspection Program O
- - - 3 I.8.2.l(1) Verify the Adequacy of Management and Procedural Controls Sege OIE/DQASIP/RCPB LI (NOTF 3) 11/30/83 NA ts and Staff Discipline I,b.2.l(2) Verify that Systems Required to Be Operable Are Properly Sege Aligned OIE/OQASIP/RCPB LI (NOTE 3) 11/30/83 NA l
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La TABLE II (Continued) c3 2g Actio, Lead Lead Office / Safety latest cm Plan Item / SPEB Division / Priority latest I s suance MPA Issue No. Title Engineer Branc h Ranking Revision Date No.
I.B 2.1(3) Follow up on Completed Maintenance work Orders to Sege OIE/DQASIP/RCPB LI (NOTE 3) 11/30/83 NA Assure Proper Testing and Return to Service I.B.2.1(4) Observe Surveillance Tests to Determine whether Test Sege OIE/DQASIP/RCPB LI (NOTE 3) 11/30/83 NA Instruments Are Properly Calibrated I.B.2.l(5) Verify that Licensees Are Complying with Technical Sege OIE/DQASIP/RCPB LI (NOTE 3) 11/30/83 NA Specifications I . B. 2. l(6 ) Observe Routine Maintenance Sege OIE/DQASIP/RCPB L1 (NOTE 3) 11/30/83 NA I . B . 2. l( 7) Inspect Terminal Boards, Panels, and Instrument Racks Sege OIE/DQASIP/RCPB LI (NOTE 3) 11/30/83 NA for Unauthorized Jumpers and Bypasses I.B.2.2 Resident Inspector at Operating Reactors Sege OIE/DQASIP/ORPB LI (NOTE 3) 11/30/83 NA I . B. 2. 3 Regional Evaluations Sege OIE/DQASIP/ORPB LI (NOTE 3) 11/30/83 NA I.B.2.4 Overview of Licensee Perf ormance Sege OIE/DQASIP/0RPB LI (NOTE 3) 11/30/83 NA
!.C OPERATING PROCED8dE5 I . C.1 Short-Ters Accident Analysis and Procedures Revision - - -
I.C 1(1) Small Break LOCAs -
NRR I I.C.l(2) Inadequate Core Cooling -
NRR I F-04 BO !.C.l(3) Transients and Accidents -
NRR I F-05 I.C.l(4) Confirmatory Analyses of Selected Transients Riggs NRR/D5I/R58 NOTE 3(b) 2 12/31/85 NA I.C.2 Shif t and Relief Turnover Procedures -
NRR I I.C.3 shift Supervisor Responsibilities -
NRR I
!.C.4 Control Room Access -
NRR I I.C.5 Procedures for Feedback of Operating Experience to -
NRR/DL I F-06 Plant Staff I.C.5 Procedures for Verification of Correct Performance of -
NRR/DL I F-07 Operating Activities I.C.7 N555 Vendor Review of Procedures -
NRR/DHF5/PSRB 1
!.C.8 Pilot Monitoring of Selected Emergency Procedures for -
NRR/DHF 5/PSRB I Near-Tern Operating License Applicants I.C.9 Long-Term Program Plan for Upgrading of Procedures Riggs NRR/DHF5/PSRB HF 01. 4. 2, 2 12/31/85 NA HF01.4.4, HF02 I;D CONTROL ROOM DF51CN 35 c: I.D.1 Control Raos Design Reviews -
NRR/DL I F-08 po b$ I.D.2 I.D.3 Plant Safety Parameter Display Console Safety System Status Monitoring Thatcher NRR/DL NRR/DHF 5/HFI B I
MEDIUM 2 12/31/85 F-09 ll
-a
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c3 I.D.4 Control Room Design Standard Thatcher NRR/DHF5/HFEB HF01.5.3 2 12/31/85 NA ((
OD I.0.5 Improved Control Room Instrumentation Research - - -
c)
U$ I.D.5(1) Operator-Process Communication Thatcher RE S/DF 0/HF BR NOTE 3(b) 2 12/31/85 NA '3 Ln O O O
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o TAElf II (Continued)
N 00 Action Lead Lead Office / Safety Latest
- Plan Item / SPE8 Division / Priority Latest issuanc e MPA Issue No. Title Engineer Branch Ranking Revision Date No.
l.D 5(2) Plant Status and Post-Accident Munitoring Thatcher RE5/DF0/HFBR NOTE 3(a) 2 12/31/85 1.D.5(3) On-Line Reactor Surveillance System Thatcher RES/DET/EEIC8 N0f f 1 2 12/31/u5 1.D.5(4) Process Monitoring Instrumentation Thatcher RES/DF0/ICBR NOTE 3(b) 2 12/31/85 NA
!.D.5(5) Disturbance Analysis Systems Thatcher NRR/DHF5/HFEB HF01.5.4 2 12/31/85 NA I.D.6 Technology Transfer Conf erence Thatcher RE S/DF 0/HF ER L1 (NOTE 3) 2 12/31/85 NA LE ANALY5[5 AND DISSEMINATION OF OPERATING EXPERIENCE I.E.1 Office for Analysis and Evaluation of Operational Matthews AE00/Pl$ LI (NOTE 3) I 6/30/84 NA Data I.E.2 Program Office Operational Data Evaluation Matthews NRR/DL/0RA8 LI (NOTE 3) 1 6/30/84 NA I.E.3 Operational Safety Data Analysis Matthews RES/DRA/kRBR LI (NOTE 3) 1 6/3U/84 NA I.E.4 Coordination of Licensee, Industry, and Regulatory Matthews AEOD/PTB L1 (NOTE 3) 1 6/30/84 NA Programs
- 1. E . 5 Nuclear Plant Reliability Data System Matthews AE00/PTB LI (NOTE 3) 1 6/30/84 NA
!.E.6 Reporting Requirements Matthews AE00/PTB Li (NOTE 3) 1 6/30/84 NA
!.E.7 Foreign Sources Matthews IP LI (NOTE 3) 1 6/30/84 NA g I.E.8 Human Error Rate Analysis Matthews RES/DF0/HFBR LI (NOTE 3) 1 6/30/84 NA ui M
QUALITY ASSURANCE I . F .1 Expand QA List Pittman 01E/DQASIP/QUA8 HIGH I 12/31/85 1.F.2 Develop More Detailed QA Criteria - - -
I.F.2(1) Assure the Independence of the Organization Performing Pittman O!E/DQA51P/QUA8 LOW l 12/31/85 NA the Checking Function I.F.2(2) Include QA Personnel in Review and Approval of Plant Pittman O!E/DQASIP/QUAB NOTE 3(a) 1 12/31/85 NA Procedures I.F.2(3) Include QA Personnel in All Design, Construction, Pittman OIE/DQA5!P/QUAB NOTE 3(a) 1 12/31/85 NA Installation. Testing, and Operation Activities I.F.2(4) Establish Criteria for Determining QA Requirements Pittman OIE/DQA5!P/QUAB LOW l 12/31/85 NA for Specific Classes of Equipment I.F.2(5) Establish Qualification Requirements for QA and QC Pittman OIE/DQASIP/QUA8 LOW I 12/31/85 NA Personnel I.F.2(6) Increase the Size of Licensees
- QA Staff Pittman OIE/DQA51P/QUAB NOTE 3(a) 1 12/31/85 NA I.F.2(7) Clarify that the QA Program Is a Condition of the Pittman OIE/DQASIP/QUAB LOW l 12/31/85 NA
- u I . F . 2(8)
Construction Permit and Operating License Compare NRC QA Requirements with Those of Other Pittman OIE/DQASIP/QUA8 LOW I 12/31/85 NA :x3
$ Agencies @
s !.F.2(9) Clarify Organizational Reporting Levels for the QA Pittman OIE/DQASIP/QUA8 NOTE 3(a) 1 12/31/85 NA -a w I.F.2(10)
Organization Clarify Requirements for Maintenance of "As-Built" Pittman LOW I 12/31/85 NA i
o 01E/DQA5]P/QUA8 W Documentation D I.F.2(ll) Define Role of QA in Design and Analys,is Activities Pittman O!E/DQASIP/QUA8 LOW l 12/31/85 NA Us
o Ch TABtf II (Continued) o Action Lead Lead Office / Safety latest CD Plan item / SPE B Division / Priority Latest Issuance MPA Issue No. Title Engineer Branch Ranking Revision Date No.
PRE OPE RATIONAL ANO LOW-POWE R TE STING I.G.1 Training Requirements -
NRR/DHF5/P5RB I I.G 2 Scope of Test Program V'Molen NRR/DHF5/PSRB NOTE 3(a) 1- 12/31/84 NA II_A SITING II A.1 Siting Policy Reformulation V'Molen NRR/DE/SAB NOTE 3(b) 1 12/31/84 NA II.A.2 Site tvaluation of Existing facilities V'Molen NRR/DE/SAB V.A.1 1 12/31/84 NA y CONSIDERATION OF DEGRADED OR MELTED CORES IN SAF Eiv REVIEW II . B.1 Reactor Coolant System Vents -
NRR/DL I F-10 11 B.2 Plant Shielding to Provide Access to Vital Areas and -
NRR/DL I F-ll Protect Safety Equipment for Post-Accident Operation ro !! .B. 3 Post-Accident 5ampling -
NRR/DL I F-12 m II.B.4 Training for Mitigating Core Damage -
NRR/DL 1 f-13
!! ii 5 Research on Phenomena Associated with Core Degradation - - -
and Fuel Melting II.B.5(1) b=Navior of Severely Damaged Fuel V'Molen RES/DAE/FSRB HIGH I 12/31/B5
!!.B 5(2) Behav br of Core Melt V'Molen RES/DAE/CSRB HIGH 1 12/31/85 II.B 5(3) Effect or Hydrogen Burning and Explosions on V'Malen RES/DAE/CSRB MEDIUM 1 12/31/85 Containment structure
!!.B.6 Risk Reduction for Operating Reactors at Sites with Pittman NRR/ DST /RRAB NOTE 3(a) 1 12/31/85 High Population Densities II.B 7 Analysis of Hydrogen Control Matthews NRR/DSI/CSB II.B.8 1 12/31/85 II.B 8 Rulemaking Proceeding on Degraded Core Accidents V'Molen RES/DRA0/RAMR NOTE 3(a) 1 12/31/85
_Il C ret! ABILITY ENGINEERING AND RISK ASSESSMENT
!!.C.1 Interim Reliability Evaluation Program Pittman RES/DRA0/RRB NOTE 3(b) 1 12/31/85 NA II.C.2 Continuation of Interim Reliability Evaluation Program Pittman NRR/ DST /RRAB NOTE 3(b) 1 12/31/85 NA 2 !!.C_3 Systems Interaction Pittman NRR/ DST /GIB A-17 1 12/31/85 NA C II C.4 Reliability Engineering Pittman RES/DRA0/RRB HIGH I 12/31/85 rn #
Q M RE ACTOR COOL ANT SYSTEM RELIEF AND SAFETY VALVES 1
o us LO
- j !!.D.1 Testing Requirements -
NRR/DL I F-14 $
!!.D.2 Research on Relief and Saf ety Valve Test Requirements Riggs RES LOW 11/30/83 NA II.D.3 Relief and Safety Valve Position Indication -
NRR I
- O O O
_ _ . . . . _ . . . _ _ . _ . _ - . _ _ _ . _ . . . _ _ . _ . _ _ _ . . .m _.. _ . . . _ __ _ . . . - -. _
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On T A8t f II (Corit inued)
I o f h 05 Action Lead Lead Office / Safety Latest 1
Plan item / SPE8 Division / Priority Latest Issuance MPA I Issue No. Title Engineer Branch Ranking Revision Date No.
J II.E SYSTEM DESIGN 4 =
' II . E .1 Auxiliary Feedwater System l I I . E .1.1 Auxiliary Feedwater System Evaluation -
NRR/DL I l-15
, II.E.1.2 Auxiliary Feedwater System Automatic Initiation and -
NRR/DL ,I f - 16, F - 17 Flow Indication l I I . E .1. 3 update Standard Review Plan and Develop Regulatory Riggs i Guide RES/DRA/RRBR NOIE 3(a) If/30/83 II.E.2 Emergency Core Cooling System I I . E . 2.1 Reliance on ECC5 Riggs l' II.E.2.2
. NRR/051/R58 II.K.3(17) 1 12/31/85 NA Research on Small Break I"CAs and Anomalous Transients Riggs RES/DAE/R5RB NOTE 3(b) 1 12/31/85 NA
, II.E.2.3 Uncertainties in Pe='orrance Predictions V'Molen NRR/051/R$8 LOW I 12/31/85 NA 4
II.E. 3 D,ecay Heat Removal i II . E . 3.1 Reliability of Power Supplies for Natural Circulation -
NRR I 4
!! . E . 3. 2 Systems Reliability V'Molen NRR/ DST /GIB A-45 11/30/83 NA II.E.3.3 Coordinated Study of Shutdown Heat Removal Requirements V'Molen NRR/ DST /GIB A-45 11/30/83 NA j U !!.E.3.4 Alternate Concepts Research Riggs RES/DAE/FBR8 NOTE 3(b) 11/30/83 NA
, ll.E.3.5 Regulatory Guide Riggs NRR/ DST /GIB A-45 11/30/83 NA -
1
.i !!.E.4 Containment Design
- II.E.4.1 Dedicated Penetrations -
NRR/DL I F-18 i II.E.4.2 Isolation Dependability -
NRR/DL I F-19
- II.E.4.3 Integrity Check Milstead NRR/D5!/C58 HIGH 11/30/83
, II.E.4.4 Purging - - -
II.E.4.4(1) Issue Letter to Licensees Requesting Limited Purging Milstead NRR/DSI/CSB NOTE 3(a) 11/30/83
- ll.E.4.4(2) Issue Letter to Licensees Requesting Information on Milstead NRR/051/C5J NOTE 3(a) 11/30/83 I
Isolation Letter II.E.* 4(3) Issue Letter to Licensees on Valve Operability Milstead NRR/DSI/CSB NOTE 3(a) 11/30/83 II.E.4.4(4) Evaluate Purging and Venting During Normal Operation Milstead NRR/DSI/CSB NOTE 3(b) 11/30/83 NA II.E.4.4(5) Issue Modified Purging and Venting Requirement Milstead NRR/051/C58 NOTE 3(b) 11/30/83 NA i
i II.E.5 Design Sensitivity of B&W Reactors
! II.E.5.1 Design Evaluation - Thatcher NRR/DSI/R58 II.E.! 2 B&W Reactor Transient Response Task Force NOTE 3(a) 1 12/31/84' Z Thatcher NRR/DL/0RAB NOTE 3(a) .1 12/31/84 l
} 5 II.E.6 In Situ Testing of Valves g
Q5 II.E.6.1 Test Adequacy Study Thatcher NRR/DE/MEB MEDIUM 11/30/83. <
- w o i w 3 s
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1 i
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C3 04
{} TA8tf II (Cont inued)
C3
%s Lead Office / Safety Latest CD Actian Lead CN Plan Item / SPEB Division / Priority Latest Issuance MPA Engineer B ranc h Ranking Revision Date No.
Issue No. Title
_II F INSTRUMENTATION AND CONTROLS Additional Accident Monitoring Instrumentation -
NRR/DL I F -20, F -21, II . F .1 F -22 F-23 F-24, F-25 II.F.2 Identification of and Recovery from Conditions -
NRR/DL I t-26 Leading to Inadequate Core Cooling
?I.F.3 Instruments for Monitoring Accident Conditions V'Molen RES/DF0/ICBR NOTE 3(3) 11/30/83 II.F.4 Study of Control and Protective Action Design Thatcher NRR/DSI/ICSB CRGr 11/30/83 NA Requirements il.F.5 Classification of Instrumentation, Control, and Thatcher RES/DET/EEICB MEDIUM 11/30/83 Electrical Equipment II.G ELECTRICAL POWER II.G.1 Power Supplies for Pressurizer Relief Valves, Block -
NRR I Valves, and Level Indicators CD ll;H TMI-2 CLEANUP AND EXAMINATION II.H.1 Maintain Safety of TMI-2 and Minimize Environmental Matthews NRR/TMIPO NOTE 3(b) 11/30/83 NA Impact II.H.2 Obtain Technical Data on the Conditions Inside the Milstead RES/DAE/FSRB HICH 11/30/83 TMI-2 Containment Structure II.H.3 Evaluate and Feed Back Information Obtained from TMI Milstead NRR/TMIPO II.H.2 11/30/83 NA II.H.4 Determine Impact of TMI on Socioeconomic and Real Milstead RES/DH5WM/SEBR LI (NOTE 3) 11/30/83 NA Property Values ll.J GENERAL IMPLICATIONS OF TMI FOR DESIGN AND CONSTRUCTION ACIIVITIES II.J.! Vendor Inspection Program II.J.l.1 Establish a Priority System for Conducting Vendor Riani OIE/DQASIP LI (NOTE 3) 11/30/83 NA mg c: Inspections Modify Existing vendor Inspection Program Riani OIE/DQASIP LI (NOTE 3) 11/30/83 NA $3 f5 II.J.1.2 Riani LI (NOTE 3) 11/30/83 NA ac C) II . J.1. 3 Increase Regulatory Control Over Present Non-Licensees OIE/DQASIP II.J.1.4 Assign Resident Inspectors to Reactor Vendors and Riani OIE/DQASIP LI (NOTE 3) 11/30/83 NA $7 c'3 p Archi tec t-E ngi neers
{D 3 00 tn 9 O O
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01
% IABLE !! (Continued)
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Action Plan Item /
Lead 5PEB LeaJ Office /
Div ica/
Safety Priority latest latest is>uante MFA Issue No. Title Engineer B ranc h Ranking Revision Date No.
!!.J.2 Construction Inspection Program -
II.J.2.1 Reorient Construction Inspection Program Riani OIE/0QA51P 'll (N0ff 3) 11/30/83 NA II.J.2.2 Increase Emphasis on Independent Measurement in Riani OIE/0QA51P L1 (NOTE 3) 11/30/83 NA Construction Inspection Program II.J.2.3 Assign Resident Inspectors to All Construction Sites Riani OIE/DQASIP Li (NOTE 3) 11/30/83 NA II.J 3 Management for Design and Construction ll.J.3.1 Organization and Staffing to Oversee Design and Pittman NRR/0HF5/LQB 1. 8.1.1 11/30/83 NA Construction
! ! . J. 3. 2 Issue Regulatory Guide Pittman NRR/0HF5/tQB 1. 8.1.1 11/30/83 NA
!!.J.4 Revise Deficiency Reporting Requirements
! ! . J. 4.1 Revise Deficiency Reporting Requirements Riani ole /DEPER/EAB NOTE 2 11/30/83
!!.K MEASURES TO MITIGATE SMALL-BREAK LOSS-OF-COOLANT ACCIDENTS AND LOSS-OF-FEEDWATER ACCIDENTS m I I . K .1 IE Bu11etins - - -
LD II . K.1(1) Review TMI-2 PNs and Detailed Chronology of the Enrit NRR NOTE 3(a) 12/31/84 -
TMI-2 Accident II.K.!(2) Review Transients Similar to IMI-2 That Have Emrit NRR NOTE 3(a) 12/31/84 -
Occurred at Other Facilities and NRC Evaluation of Davis-Besse Event II.K.I(3) Review Operating Procedures for Recognizing, Eerit NRR NOTE 3(a) 12/31/84 -
Preventing, and Mitigating Void Formation in Transients and Accidents ll.K.1(4) Revitw Operating Procedures and Training Eerit NRR NOTE 3(a) 12/31/84 -
Instructions ll.K.l(5) Safety-Related Valve Position Description Enrit NRR NOTE 3(a) 12/31/84 -
!!.K.l(6) Review Containment isolation Initiation Design Eerit NRR NOTE 3(a) 12/31/84 -
and Procedures II.K.l(7) Implement Positive Position Controls on Valves Emrit NRR NOTE 3(a) 12/31/84 -
That Could Compromise or Defeat AFW Flow
!!.K.l(8) Implement Procedures That Assure Two Independent Emrit NRR NOTE 3(a) 12/31/84 -
100% AFW Flow Paths ll K. l(9) Review Procedures to Assure That Radioactive Emrit NRR NOTE 3(a) 12/31/84 -
h A
Liquids and Gases Are Not Transferred out of g
Containment Inadvertently to
$ e II.K.l(10) Review and Modify Procedures for Removing Safety-Related Systems from Service Emrit NRR NOTE 3(a) 12/31/84 -
1 em
(*)
- 11. K. l(II) Make All Operating and Maintenance Personnel Aware of the Seriousness and Consequences of the Emrit NRR NOTE 3(a) 12/31/84 -
p
s E rroneous Actions leading up to, and in Early ,
Phases of, the TMI-2 Accident
CD 04
{l; TABLE II (Continued) c)
%s 00 Actio) Lead lead Office / Safety latest C' Plan item / SPEB Division / Priority Latest Is suance MPA Issue No. Title Engineer Branch Ranking Revision Date No.
!!.K.l(12) One Hour Notification Requirement and Continuous Emrit NRR NOTE 3(a) 12/31/84 -
Communications Channels II . K. l(13) Propose Technical Specification Changes Reflecting Emrit NRR NOTE 3(a) 12/31/84 -
Implementation of All Bulletin Items
!!.K.1(14) Review Operating Modes and Procedures to Deal with Emrit NRR NOTE 3(a) 12/31/84 -
Significant Amounts of Hydrogen
!!.K.l(15) for Facilities with Non-Automatic AFW Initiation, Enrit NRR NOTE 3(a) 12/31/84 -
Provide Dedicated Operator in Continuous Communication with CR to Dperate AfW II.K.l(16) Implement Procedures That Identify PRZ PORV "Open" Emrit NRR NOTE 3(a) 12/31/84 -
Indications and That Direct Operator to Close Manually at " Reset" Setpoint II.K.l(17) Trip PZR Level Histable so That PZR Low Pressure Enrit NRR N0fE 3(a) 12/31/84 -
Will Initiate Safety Injection I I . K. l( 18 ) Develop Procedures and Train Operators on Methods Emrit NRR N01E 3(a) 12/31/84 -
of Establishing and Maintaining Natural Circulation ll.K.l(19) Describe Design and Procedure Modifications to Emrit NRR NOTE 3(a) 12/31/84 -
Reduce Likelihood of Automatic PZR PORV Actuation in Transients
$$ !!.K.l(20) Provide Procedures and Training to Operators for Eerit NRR NOTE 3(a) 12/31/84 -
Prompt Manual Reactor Trip for LOFW, TT, MSIV Closure, LOOP, LO5G Level, and LO PIR Level II.K.l(21) Provide Automatic Safety-Grade Anticipatory Reactor Emrit NRR NOTE 3(a) 12/31/84 -
Trip for LOFW, TT. or Significant Decrease in SG Level II.K.l(22) Describe Automatic and Manual Actions for Proper Emrit NRR NOTE 3(a) 12/31/84 -
Functioning of Auxiliary Heat Removal Systems When fW System Not Operable II.K.1(23) Describe uses and Types of RV Level Indication for Emrit NRR NOTE 3(a) 12/31/84 -
Automatic and Manual Initiation Safety Systems II.K.l(24) Perform LOCA Analyses for a Range of Small-Break Emrit NRR NOTE 3(a) 12/31/84 -
Sizes and a Range of Time lapses Between Reactor Trip and RCP Trip II.K.l(25) Develop Operator Action Guidelines Emrit NRR NOTE 3(a) 12/31/84 -
II.K.l(26) Revise Emergency Procedures and Train R0s and SR0s Emrit NRR NOTE 3(a) 12/31/84 -
II.K.l(27) Provide Analyses and Develop Guidelines and Emrit NRR NOTE 3(a) 12/31/84 -
Procedures for Inadequate Core Cooling Conditions j{ ll.K.l(26) Provide Design That Will Assure Automatic RCP Trip Emrit NRR NOTE 3(a) 12/31/84 -
pg for All Circumstances Where Required as
[]
s II . K. 2 II.K.2(1)
Commission Orders on B&W Plants Upgrade Timeliness and Reliability of AFW System Emrit NRR/051 NOTE 3(a) !?/31/84 -
((
-u C$ ll.K.2(2) Procedures and Training to Initiate and Control Emrit NRR NOTE 3(a) 12/31/84 -
((
og AFW Independent of Integrated Control System c)
OJ !!.K 2(3) :3 Hard-Wired Control-Grade Anticipatory Reactor Trips Emrit NRR/D51 NOTE 3(a) 12/31/84 -
II.K.2(4) Small-Break LOCA Analysis, Procedures and Operator Emrit NRR/DHF 5/0LB N0il 3(a) 12/31/84 -
um Training O O O
[
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cn g 1ABLF II (Contined) k 00
~
Action Lead Lead Office / Safety latest CD Plan Item / SPEB Divisiun/ Priority latest Issuance MPA Issue No. Title Engireer B ranc h Ranking Revision Date No.
II . K. 2( 5 ) Complete TMI-2 Simulator Training for All Operators Emrit NRR NOTE 3(a) 12/31/84 -
II.K.2(6) Reevaluate Analysis for Dual-Level Setpoint Control Emrit NRR/DSI NOTE 3(a) 12/31/84 -
II.K.2(7) Reevaluate Transient c.f September 24, 1977 Emrit NRR/051 NOIE 3(.a) 12/31/84 -
II.K.2(8) Continued Ungrading of AFW System Emrit NRR II.E.1.1, 12/31/84 NA II.E.1.2 II.K.2(9) Analysis and Upgradir.g of Integrated Control System Emrit NRR I 12/31/84 F-27 II.K.2(10) Hard-Wired Safety-Grade Anticipatory Reactor Trips Emrit NRR I 12/31/84 F-28 i !!.K.2(ll) Operator Training and Drilling Emrit NRR I 12/31/84 F-29
!! K.2(12) Transient Analysis and Procedures for Management Emrit NRR I.C.l(3) 12/31/84 NA of Small Breaks I I . K. 2( 13 ) Thermal-Mechanical Report on Ef fect of HPI on Vessel Emrit NRR I 12/31/84 F-30 Integrity for Small-Break LOCA With No AFW
!!.K.2(14) Demonstrate That Predicted Lift Frequency of PORVs Emrit NRR I 12/31/84 F-31 and SVs Is Acceptable
!!.K 2(15) Analysis of Ef fects of Slug Flow on Once-Through Emrit NR9 I '12/31/84 -
Steam Generator Tubes After Primary System Voiding
!!.K.2(16) Impact of RCP Seal Damage following Small-Break Emrit NRR I 12/31/84 F-32 LOCA With Loss of Of fsite Power II.K 2(17) Analysis of Potential Voiding in RCS During Emrit NRR I 12/31/84 F-33 g Anticipated Transients II.K.2(18) Analysis of loss of Feedwater and Other Anticipated Emrit NRR I.C.l(3) 12/31/84 NA Transients II.K.2(19) Benchmark Analysis of Sequential AFW Flow to Once- Emrit NRR I 12/31/84 F-34 Through 5 team Generator II.K.2(20) Analysis of Steam Response to small-Break LOCA Eerit NRR I 12/31/84 F-35 That Causes System Pressure to Exceed PORV Setpoint II.K.2(21) LOFT L3-1 Predictions Emrit NRR/DSI NOTE 3(a) 12/31/84 -
II K.3 Final Recommendations of Bulletins and Orders Task - - -
Force II . K. 3( 1) Install Automatic PORV Isolation 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 Emrit NRR I 12/31/84 F-37 System II.K.3(3) Report , Safety and Relief Valve Failures Promptly Emrit NRR I 12/31/84 F-38 and Challenges Annually II. K. 3(4) Review and Upgrade Reliability and Redundancy of Emrit NRR II.C.1, 12/31/84 NA Non-Safety Equipment for Small-Break LOCA Mitigation II.C.2, m !!.K.3(5) Automatic Trip of Reactor Coolant Pumps Emrit II.C.3 NRR .I 12/31/84 F-39, G-01 m II.K.3(6) Instrumentation to Verify Natural Circulation Emrit NRR/DSI I.C.l(3), 12/31/84 II.F.2, NA w II.K.3(7) Evaluation of PORV Opening Probability During II.F.3 W Emrit NRR ! 12/31/84 -
o Overpressure Transient 3 t!i
C3 CN
{]; TK8tf II (Continued)
C3 00 Action Lead Lead Office / Safety Latest O' Plan Item / SPE8 Division / Friority latest Issuance MPA Issue No. Title Engineer Branch Ranking Revision Date No.
II.K.3(8) Further Staf f Consideratiors of Need for Diverse Emrit NRR/ DST /GIB II.C.1, 12/31/84 NA Decay Heat Removal Method Independent of SGs II.E.3.3 II.K.3(9) Proportional Integrat Derivative Controller Emrit NRR I 12/31/84 F-40 Modification II.K.3(10) Anticipatory Trip Modification Proposed by some Emrit NRR I 12/31/84 F-41 Licensees to Confine Range of Use to High Power Levels II . K. 3(ll) Control Use of PORV Supplied by Control Components, Emrit NRR I 12/31/84 -
Inc. Until Further Review Complete II.K.3(12) Confirm Existence of Anticipatory Trip upon Turbine Eerit NRR I 12/31/84 F-42 Trip II.K.3(13) Separation of HPCI and RCIC System Initiation Levels Emrit NRR I 12/31/84 F-43 II.K.3(14) Isolation of Isolation Condensers on High Radiation Eerit NRR I 12/31/84 F-44
!!.K.3(15) Modify Break Detection Logic to Prevent Spurious Enrit NRR I 12/31/84 F-45 Isolation of HPCI and RCIC Systems II.K.3(16) Reduction of Challenges and Failures of Relief Emrit NRR I 12/31/84 F-4b Valves - Feasibility Study and System Modification
!!.K.3(17) Report on Outage of ECC Systees - Licensee Report Enrit NRR I 12/31/84 F-47 g, and Technical Specification Changes na II.K.3(18) Modification of ADS Logic Feasibility Study and Enrit NRR I 12/31/84 F-48 Modification for Increased Diversity for Some Event sequences II . K. 3( 19) Interlock on Recirculatten Pump Loops Emrit NRR I 12/31/84 F-49 II . K. 3(20) loss of Service Water for Big dock 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 Modification II.K.3(22) Automatic Switchover of RCIC System suction - Larit NRR I 12/31/84 F-51 Verify Procedures and Modify Design ll.K.3(23) Central Water Level Recording Emrit NRR I.D.2, 12/31/84 NA Ill.A.I.2(1),
Ill.A.3.4
!!.K.3(24) Confirm Adequacy of Space Cooling for HPCI and Enrit NRR I 12/31/84 F-52 RCIC Systems II.K.3(25) Effect of Loss of AC Power on Pump Seals Enrit NRR I 12/31/84 F-53
!!.K.3(26) Study Effect on RHR Reliability of Its Use for Emrit NRR/DSI II.E.2.1 12/31/84 NA Fuel Pool Cooling II.K.3(27) Provide Common Reference Level for Vessel Level Emrit NRR I 12/31/84 F-54
[E Instrumentation 30 II . K. 3(28) Study and Verify Qualification of Accumulators Emrit NRR I 12/31/84 F-55 33 E) on ADS Valves [!
s II.K.3(29) Study to Demonstrate Performance of Isolation Emrit NRR I 12/31/84 F-56 -a
$$ Condensers with Non-Condensibles El ca !!.K.3(30) Revised Small-Break LOCA Methods to Show Compliance Emrit NRR I 12/31/84 F-57 c)
'# with 10 CF R 50, Appendix K 3 II.K 3(31) Plant-Specific Calculations to Show Compliance with Emrit NRR 1 12/31/84 F-58 tn 10 CFR 50.46 O O O
m _ _ - . . _ _ _ __ _m_-___ --- -_ __ ___ _~ _-. . . _ - _ m._ _ . _ .=_- ---__ -- _- . ._ _ ._
1 1
i 1
i o m
,' N TABLE II (Continued) i w o
, h m
Action Plan Item /
Division /
Safety Priority Latest latest Issuance MPA 4
l Issue No. Title Engineer Branch Ranking Revision Date No.
l II.K.3(32) Provide Experimental Verification of Two-Phase Eerit NRR/DSI II.E.2.2 12/31/84 NA'
, Natural Circulation Models j II. K. 3( 33) Evaluate Elimination of PORV Function Eerit NRR II.C.1 12/31/84 NA a II.K.3(34) Relap-4 Model Development Enrit NRR/DSI II.E.2.2 12/31/84 NA ,
i II.K.3(35) Evaluation of Ef fects of Core Flood Tank Injection Enrit NRR I,C.l(3) 12/31/84 NA !
j on Small-Break LOCAs 4
I I . K. 3( 36 ) Additional Staff Audit Calculations of B&W Small- Enrit NRR I.C.1 12/31/84 NA i Break LOCA Analyses
} II. K. 3( 37) Analysis of B&W Response to Isolated Small-Break Emrit NRR I.C.l(3) 12/31/84 NA 1 LOCA
] II. K. 3(38) Analysis of Plant Response to a Small-Break LOCA in Emrit NRR I.C.1(3) 12/31/84 NA j the Pressurizer Spray Line ;
1 II . K. 3( 39) Evaluation of Effects of Water Slugs in Piping Enrit NRR I.C.l(3) 12/31/84 NA Caused by HPI and CFT Flows j II.K.3(40) Evaluation of RCP Seal Damage and Leakage During Enrit NRR II.K.2(16) 12/31/84 NA l
4 Small-Break LOCA J II . K. 3(41) Submit Predictions for LOFT Test L3-6 with RCPs Emrit NRR I.C.l(3) 12/31/84 NA l Running II. K. 3(42) Submit Requested Information on the Effects of Enrit NRR I.C.1(3) 12/31/84 NA t
d II.K.3(43)
Non-Condensible Cases Evaluation of Mechanical Effects of Slug Flow on Enrit NRR .II.K,2(15) 12/31/84 NA I
j Steam Generator Tubes j II.K.3(44) Evaluation of Anticipated Transients with Single Eerit NRR I 12/31/84 F-59 Failure to Verify No Significant Fuel Failure
- II.K.3(45) Evaluate Depressurization with Other Than Full ADS Enrit NRR I 12/31/84 F-60 II.K.3(46) Response to List of Concerns from ACRS Consultant Enrit NRR I 12/31/84 F-61
] II.K.3(47) Test Program for Small-8reak LOCA Model Verification Enrit NRR I.C.1(3), 12/31/84 NA j Pretest Prediction Test Program, and Model II.E.2.2 i Verification i II.K.3(48) Assess Change in Safety Reliability as a Result of Eerit NRR II.C.1, 12/31/84 NA Implementing B&OTF Recommendations II.C.2 i II.K 3(49) Review of Procedures (NRC) Enrit NRR/DHFS/PSRB I.C.8, 12/31/84 NA j I.C.9 II.K.3(50) Review of Procedures (NS55 Vendors) Enrit NRR/DHFS/PSR8 I.C.7, 12/31/84 NA l I.C.9 j II.K.3(51) Symptom-Based Emergency Procedures Enrit NRR/DHFS/PSR8 I.C.9 12/31/84 NA
{ II.K 3(52) Operator Awareness of Revised Emergency Procedures Enrit NRR I . B. I.1, 12/31/84 NA j g 1.C.2,
, ao I.C.5 30 i g II. K. 3(53)
II.K.3(54)
Two Operators in Control Room Eerit NRR I.A.I.3 12/31/84 NA $
i e Simulator upgrade for Small-Break LOCAs Enrit NRR I.A.4.l(2) 12/31/84 NA j o II.K 3(55) Operator Monitoring of Control Board Enrit NRR I.C.1(3), 12/31/84 NA E j w I.D.2, o j W I.D.3 3 II.E.3(56) Simulator Training Requirements Enrit NRR/DHFS/0LB !.A.2.6(3), 12/31/84 NA un I.A.3.1 i II.K.3(57) Identify Water Sources Prior to Manual Activation Emrit NRR I 12/31/84 F-62 [
i of ADS '
I i
f, _
c)
Va TAME !! (Continued) k 00 Action Lead Lead Office / Safety Latest CN Plan item / SPEB Division / Priority latest I s suanc e MPA Issue No. Title Engineer B ranc h ' Ranking Revision Date No.
III.A EMERGENCY PPEPAREDNESS AND RADI ATION EFF ECTS III.A.1 Improve Licensee Emergency Preparedness - Short Tern Ill.A 1.1 Upgrade Emergency Preparedress - - -
!!!.A.I.l(1) Implement Action Plan Requirements for Promptly -
OIE/DEPER/EPB I Improving Licensee Emergency Preparedness
!!! A.l.l(2) Perform an Integrated Assessment of the Implementation -
OIE/DEPER/EPB !
!!I . A I. 2 Upgrade Licensee Emergency Support Facilities - - -
III.A 1.2(1) Technical Support Center -
OIE/DEPER/EPB I f-63 Ill.A.I.2(2) On-Site Operational Support Center -
OIE/DEPER/EPB 1 F-64
!!!.A.I.2(3) Near-Site Emergency Operations f acility -
OIE/DEPER/EPB I l-65 111.1 1.3 Maintain Supplies of Thyroid-Blocking Agent - - -
!!!. A. I. 3(1) Workers Riggs OIE/DEPER/EPB NOTE 3(b) 1 12/31/85 NA
!!! . A 1. 3(2) Public Riggs OIE/DEPER/EPB NOTE 3(b) 1 12/31/85 NA III.A.2 Improving Licensee Emergency Preparedness-long Term Ill.A.2.1 Amend 10 CFR 50 and 10 CFR 50, Appendix E - - -
!!!.A.2.l(l) Publish Proposed Amendments to the Rules -
RES I jj !!I.A.2.l(2) Conduct Public Regional Meetings -
RES I
!!I.A.2.l(3) Prepare Final Commission Paper Recommending Adoption -
RES I of Rules III . A. 2. l(4 ) Revise Inspection Program to Cover Upgraded -
OIE I F-67 Requirements
!!!.4.2.2 Development of Guidance and Criteria -
NRR/DL I F-68
!!I. A 3 Improving NRC Emergency Preparedness
!!!.4.3.1 NRC Role in Responding to Nuclear Emergencies - - -
111.4.3.1(1) Define NRC Role in Emergency situations Riggs OIE/DEPER/IRDB NOTE 3(b) 1 6/30/85 NA 111.4.3.l(2) Revise and Upgrade Plans and Procedures for the NRC Riggs OIE/DEPER/lRDB NOTE 3(b) 1 6/30/85 NA Emergency Operations Center III.4.3.1(3) Revise Manual Chapter 0502, Other Agency Procedures, Riggs OIE/DEPER/IRDB NOTE 3(b) 1 6/30/85 NA and NUREG-0610
!!! . 4. 3. l( 4 ) Prepare Commission Paper Riggs OIE/DEPER/IRDB NOTE 3(b) 1 6/30/85 NA 111.4.3.1(5) Revise Implementing Procedures and Instructions for Riggs OIE/DEPER/lRDB NOTE 3(b) 1 6/30/85 NA Regional Offices III.A 3.2 Improve Operations Centers Riggs OIE/DEPER/lRDB NOTE 3(b) 1 6/30/85 NA
- 111.4.3.3 Communications - - -
$5 111.4.3.3(1) Install Direct Dedicated Telephone Lines Pittman OIE /DE PE R/IRDB NOTE 3(a) 1 6/30/85 NA po gg 111.4.3.3(2) Obtain Dedicated, Short-Range Radio Communication Pittman OIE/DE PE R/IRDB NOTE 3(a) 6/30/85 NA
, Systems 1
((
CD ! !! . 4. 3. 4 Nuclear Data Link Thatcher OIE/DEPER/lRDB NOTE 3(b) I b/30/85 us
[$ ! ! ! . 4. 3. 5 Iraining, Drills, and Tests Pittman OIE /DE PE R/lRDB NOTE 3(b) 1 6/30/85 NA c' La 111.4.3.6 Interaction of NRC and Other Agencies - - -
- 3 111.4.3.6(1) International Pittman OIE/DEPER/EPLB NOTE 3(b) ! 6/30/85 NA on
!!!. 4 3. 6( 2) Federal Pittman ole /DE PE R/E PL 8 NOTE 3(b) 1 6/30/85 NA III.4.3 b(3) State and Local Pittman OIE /DE PE R/E PL B N0lt 3(b) 1 6/30/85 NA O O O
m f .
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m D
C)
TA81t II (Continued)
N OD Action Lead lead Of f ice / Safety Latest
- Plan Item / SPIB Division / Priority Latest Issuance MPA Issue No. Title Engineer B ranch Ranking Revision Date No.
111.11 tME RGENCY PREPAREDNESS OF STATE ANO LOCAL GOVERNMENTS I I I . tt.1 Transfer of Responsibilities to FEMA Milstead OIE/DEPER/IRDB NOTE 3(b) II/30/H1 NA III.B.2 Implementation of NRC and FEMA Responsibilities - - -
III.B.2(1) Ihe Licensing Process Milstead OIE/DEPER/IRD8 N01E 3(b) 11/30/83 ,NA Ilf 0.2(2) federal Guidance Milstead OIE/DE PE R/IRDB NOTE 3(b) !!/30/83 NA ItI.C PUBLIC l'NFORMATION III.C.1 Have Information Available for the News Media and the - - -
Public I I I . C .1( 1) Review Publicly Available Documents Pittman PA LI (NOIE 3) 11/30/83 NA III .C . l(2) Recommend Publication of Additional Information Pittman PA LI (NOTE 3) 11/30/83 NA II I . C.. l( 3) Program of Seminars for News Media Personnel Pittman PA L1 (NOTE 3) 11/30/83 NA III.C.2 Develop Policy and Provide Training for Interfacing - - -
With the News Media g III.C.2(1) Develop Policy and Procedures for Dealing With Briefing Pittman PA LI (NOTE 3) 11/30/83 NA un Requests III.C.2(2) Provide Training for Members of the Technical Staff Pittman PA LI (NOTE 3) !!/30/83 NA III.D RA0!ATION PROTECTION
! 111.D.1 Radiation Source Control
! III.D.I.! Primary Coolant Sources Outside the Containment - - -
Structure III . D. I . l( l) Review Information Submitted by Licensees Pertaining -
NRR I l to Reducing Leakage from Operating Systems
!!I.D.I.l(2) Review Information on Provisions for Leak Detection Enrit NRR/DSI/METB NOTE 4 111. D.1.1( 3 ) Develop Proposed System Acceptance Criteria Eerit NRR/DSI/METB NOTE 4 III.D.I.2 Radioactive Gas Management Enrit NRR/051/ MET 8 DROP 11/30/83 NA 111. D.1. 3 Ventilation System and Radiolodine Adsorber Criteria - - -
!!I.D.I.3(1) Decide-whether Licensees should Perform Studies and Enrit NRR/DSI/MElB DROP 11/30/83 NA Make Modifications III.D.I.3(2) Review and Revise SRP Eerit NRR/DSI/METB DROP' 11/30/83 NA
- m 111.D.1.3(3)
I!!. D. I. 3(4 )
Require Licensees to lipgrade Filtration Systems Sponsor Studies to Evaluate Charcoal Adsorber Enrit 'NRR/D51/METB DROP 11/30/83 NA Enrit NRR/DSI/METB NOTE 3(b) 11/30/83 NA' 30
$ III.D.I.4 Radwaste System Design Features to Aid in Accident Recovery and Decontamination' Enrit NRR/D5I/METB DNOP 11/30/83 NA $
4 to to a Ill.D.2 Public Radiation Protection Improvement W o III.D.2.1 Radiological Monitoring of Effluents - - - 3 I II . D. 2. l(l) Evaluate the Feasibility and Perform a Value-Impact Eerit NRR/D5I/METB LOW 2 12/31/85 NA Lps Analysis of Modifying Effluent-Monitoring Design Criteria
o m
B LE II (Continued) o h
CD Action Plan Item /
$PEB Lead Office /
Division /
Safety Priority Latest Latest I s suanc e MPA Issue No. Title Engineer B ranc h Ranking Revision Date No.
III.0,2.l(2) Study the Feasibility of Requiring the Development Eerit NRR/D51/METB LOW 2 12/31/85 NA of Effective Means for Monitoring and Sampling Noble Cases and Radiciodine Released to the Atmosphere III.D.2.!(3) Revise Regulatory Guides Eerit NRR/051/METB LOW 2 12/31/85 NA
!!!.D.2.2 Radioiodine, Carbon-14, and Tritium Pathway Dose - - -
Analysis 1 : 1. D. 2. 2( 1 ) Perform Study of Radioiodine, Carbon-14. and Tritium Enrit NRR/D51/RAB NOTE 3(b) 2 12/31/85 NA Behavior
!!!.D.2.2(2) Evaluate Data Collected at Quad Cities Eerit NRR/051/RAB lit.D.2.5 2 12/31/85 NA I ll . D . 2. 2( 3 ) Determine the Distribution of the Chemical Species of Emrit NRR/051/RAB III.D.2.5 2 12/31/85 NA Radioiodine in Air-Water-Steam Mixtures Ill.D.2.2(4) Revise SRP and Regulatory Guides Emrit NRR/D51/RAB 111. D. 2. 5 2 12/31'/85 NA lit.D.2.3 Liquid Pathway Radiological Control - - -
111. D. 2. 3( 1) Develop Procedures to Discriminate Between Emrit NRR/DE/EHEB NOTE 3(b) 2 12/31/85 NA Sites / Plants
! ! ! . D. 2. 3(2 ) Discriminate Between Sites and Plants That Require Emrit NRR/DE/EHEB NOTE 3(b) 2 12/31/85 NA Consideration of Liquid Pathway Interdiction Techniques
!!!.D.2.3(3) Establish Feasible Method of Pathway Interdiction Emrit NRR/DE/EHEB NOTE 3(b) 2 12/31/85 NA III.D.2.3(4) Prepare a Summary Assessment Emrit NRR/DE/EHEB NOTE 3(b) 2 12/31/85 NA m Ill.D.2.4 Offsite Dose Measurements - - -
111.D.2.4(1) Study feasibility of Environmental Monitors V'Molen NRR/DSI/RAB NOTE 3(b) 2 12/31/85 NA Ill.D.2.4(2) Place 50 TLDs Around Each Site V'Molen OIE/DRP/ORPB L1 (NOTE 3) 2 12/31/85 NA
!!!.D.2.5 Offsite Dose Calculation Manual V'Molen NRR/051/RA8 NOTE 3(b) 2 12/31/85 NA
! !! . D. 2. 6 Independent Radiological Measurements V'Molen O!E/DRP/ORPB LI (NOTE 3) 2 12/31/85 NA Ill.D. 3 Worker Radiation Protection Improvement til.D.3.1 Radiation Protection Plans V'Molen NRR/051/RAB NOTE 3(b) 1 06/30/86 NA 111.D.3.2 Health Physics leprovements - - -
I ! ! . D _3. 2( 1 ) Amend 10 CFR 20 V'Molen RES/DF0/0RPBR LI (NOTE 2) 11/30/83 NA III.D.3.2(2) Issue a Regulatory Guide V'Molen RES/DF0/ORPBR LI (NOTE 3) 11/30/83 NA
!!! . D. 3. 2( 3 ) Develop Standard Performance Criteria V'Molen RES/DF0/ORPBR L1 (NOTE 2) 11/30/83 NA 111.D.3.2(4) Develop Method for Testing and Certifying Air-Purifying V'Molen RES/DF0/ORPBR Li (NOTE 2) 11/30/83 NA Respirators
!!I.D.3.3 In plant Radiation Monitoring - - -
111.D.3.3(1) Issue Letter Requiring Improved Radiation Sampling -
NRR/DL I F-69 Instrumentation III.D.3.3(2) Set Criteria Requiring Licensees to Evaluate Need for -
NRR I
'Z Additional Survey Equipment E 111. D. 3. 3( 3) Issue a Rule Change Providing Acceptable Methods for -
RE5 1 :o g Calibration of Radiation-Monitoring Instruments @
. III.D.3.3(4) Issue a Regulatory Guide -
RES I -d-O 111.0 3.4 Control Room Habitability -
NRR/DL 1 F-10 1 8
w Ill.D.3.5 Radiation Worker Exposure - - -
o 3
11! . D. 3. 5(1) Develop Format for Data To Be Collected by Utilities V'Molen RES/DF0/0RPBR LI (NOTE 5) 11/30/83 NA Regarding Total Radiation Exposure to Workers on 111.D.3.5(2) Investigative Methods of Obtaining Employee Health V'Molen RES/DF0/0RPBR L1 (NOTE 3) 11/30/83 NA Data by Nonlegislative Means Ill.D.3.5(3) Revise 10 CfR 20 V'Holen RES/Df0/ORPBR Li (NOTE 3) 11/30/83 NA O O O
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o CD TAH6E !! (Cr,ntinued) o _ _ _ . _
Action Lead Lead Of f ice / 5afety L .s t rs t CD Plan item / $PEB Division / Priority latest issuame MPA
!ssue No. Iltle Engineer Branch R.snking Revision Date No.
IV.A STRENGTHE N ENFORCEMENT PROCESS l IV.A 1 Seek Legislative Aathurity Eerit GC Li (N01E 3) 11/30/83 NA l IV.A 2 Revise Enf orcement Policy Emrit OIE/E5 L1 (NOIE 3) 11/30/83 NA l 1v. 8 1htJANCE OF INSTRUCTIONS AND INFORMAT10N TO LICENSEE 5 l
IV. H.1 Revise Practices for issuance of Instructions and Enrit OIE/DEPER L1 (NOIE 3) !!/30/83 NA I
Information to Licensees IV EXTEND LESSONS LEARNED TO LICENSED ACTIVITIES OT!!ER THAN POwE R RE ACTORS
{
IV.C.1 Extend lessons Learned from TM1 to Other NRC Programs Emrit NM55/WM NOTE 3(b) 11/30/83 NA i
w IV. D NRC STAF F TRAINING
, M X ~
l
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IV. D. I NRC Staff Training Eerit ADM/ MOTS LI (NOTE 3) 11/30/83 NA
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IV.E SAFETY DEcl510N-MAKING IV. E. ! Expand Research on Quantification of Safety Colmar RES/DRA/RABR LI (NOTE 5) 1 12/31/85 NA Decision-Making IV.E.2 Plan for Early Resolution of Safety Issues Enrit NRR/ DST /SPEB LI (NOTE 3) 1 12/31/85 NA IV. E. 3 Plan for Resolving issues at the CP Stage Colmar RES/DRA/RABR L1 (NOTE 2) 1 12/31/85 NA 2
IV.E.4 Resolve Generic Issues by Rulemaking Colmar RES/DRA/RABR LI (NOTE 5) 1 12/31/85 NA l
I I V . E . '> Assess Currently Operating Reactors Matthews NRR/DL/SEPB NOTE 3(b) 1 !?/31/85 NA IV F FINANCIAL DISINCENTIVES TO SAFETY l
z IV f .1 increased OIE Scrutiny of the Power-Ascension Test Thatcher OIE/DQA51P NOIE 3(b) 11/30/83 NA c Program O
IV f.2 Evaluate the Impacts of Financial Disincentives to the Safety of Nuclear Power Plants Matthews SP NOTE 3(b) 11/30/83 NA [
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o Cn D
o TABLE II (Continued)
N CD Actiun Lead Lead Office / Safety latest Plan Item / SPEB Division' Priority latest issuance MPA
!ssue No. Title Engineer B ranch Ranking 6evision Date No.
IV,G IMPROVE SAFETY RULEMAKING PROCEDURE S kV.G.1 Develop a Public Agenda for Rulemaking Eerit ADM/RPB L1 (NOTE 3) 11/30/83 NA IV.G.2 Periodic and Systematic Reevaluation of Existing Rules Milstead RES/DRA/RABR L1 (NOTE 5) 11/30/83 NA IV.C. 3 Improve Rulemaking Procedures Milstead RES/DR. T% LI (NOTE 3) 11/30/83 NA IV.G.4 Study Alternatives for improved Rulemaking Process Nilstead RES/LLA/kA ' L1 (NOTE 3) 11/30/83 NA
_I V . H NRC PARTICIPATION IN THE RADIATION POLICY COUNCIL IV. H.1 NRC Participation in the Radiation Policy Council Sege RES/DH5WM/HEBR L1 (NOTE 3) 11/30/83 NA TASK ACTION PLAN ITEMS A-1 Water Hammer Emrit NRN/ DST /GIB USI [ NOTE 3(a)) 1 6/30/85 NA y A-2 Asymmetric Blowdown Loads on Reactor Primary Coolant Systems Enrit NRR/ DST /GIB USI [ NOTE 3(a)] 1 6/30/85 0-10
~
A-3 Westinghouse Steam Generator Tube Integrity NRR/ DST /GIB USI 11/30/83 A-4 CE 5 team Generator Tube Integrity -
NRR/ DST /GIB USI 11/30/83 A-5 B&W Steam Generator Tube Integrity -
NRR/ DST /GIB USI 11/30/83 A-6 Mark i Short-Term Program Enrit NRR/ DST /GIB U51 [ NOTE 3(a)] 1 6/30/85 A-7 Mark I Long-Term Program Emrit NRR/ DST /GIB USI [ NOTE 3(a)] 1 6/30/85 D-01 A-8 Mark 11 Containment Pool Dyanmic Loads Long-Term Emrit NRR/ DST /GIB USI [ NOTE 3(a)] 1 6/30/85 NA Program A-9 ATWS Emrit NRR/ DST /GIB USI [ NOTE 3(a)) 1 6/30/85 A-10 BWR Feedwater Nozzle Cracking Emrit NRR/ DST /GIB USI [ NOTE 3(a)) 1 6/30/85 B-25 A-11 Reactor Vessel Materials Toughness Emrit NRR/ DST /GIB U51 [ NOTE 3(a)) 1 6/30/85 A-12 Fracture Toughness of Steam Generator and Reactor Emrit NRR/ DST /GIB USI [ NOTE 2] 1 6/30/85 NA Coolant Pump Supports A-13 $nubber Operability Assurance Emrit NRR/DE/MEB NOTE 3(a) 11/30/83 A-14 Flaw Detection Matthews NRR/DE/MTEB DROP !!/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 35 A-16 Steam Effects on BWR Core Spray Distribution Emrit NRR/DSI/CPB NOTE 3(a) 11/30/83 D-12 C A-Il Systems Interaction -
NRR/ DST /GIE: U51 11/30/83
$ A-18 Pipe Rupture Design Criteria Emrit NRR/DE /ME B DROP 11/30/83 NA [
A-19 Digital Computer Protection System Thatcher NRR/D51/IC58 NOTE 4 !!/30/83 <
o W
A-20 A-21 Impacts of the Coal f uel Cycle Main Steamline Break Inside Containment - Evaluation of V'Molen NRR/Di/EHEB NRR/051/C5B Li (NOIE 5)
LOW 11/30/83
!!/30/H3 NA NA 7
-a.
$ Environmental Conditions f or Equipment Qualification $
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g TABtE II (Continued) o Action Lead Lead Office / Safety
- tatest Plan Item / SPEB Division / Priority tatest issuance MPA 155ue No. Title Engineer Branch Ranking Revision Date No.
A-22 PWR Main Steamline Break - Core, Reactor Vessel and V'Molen NRR/D5I/C5B DROP 11/30/83 NA Containment Building Response A-23 Containment Leak Testing Matthews NRR/D51/C5B 11/30/83 A-24 RI (NOTE 5)
Qualification of Class IE Safety-Related Equipment -
NRR/ DST /GIB USI [ NOTE 3(a)] 1 6/30/85 B-60 A-25 Non-Safety Loads on Class IE Power Sources Thatcher - NRR/DSI/PSB NOTE 3(a) 11/30/83 A-26 Reactor Vessel Pressure Transient Protection -
NRR/ DST /GIB 6/30/85 B-04 USI [ NOTE 3(a)] 1 l A-27 Reload Applications -
NRR/DSI/CPB LI (NOIE 5) 11/30/83 NA i A-2B Increase in Spent Fuel Pool Storage Capacity Colmar NRR/DE/5GEB 11/30/83 NOTE 3(a)
A-29 Nuclear Power Plant Design for the Reduction of Colmar NRR/DSI/ASB MEDIUM 11/30/83 Vulnerability to Industrial Sabotage A-30 Adequacy of Safety-Related DC Power Supplies Sege NRR/DSI/PSB A,31 HIGH 11/30/83 RHR $hutdown Requirements -
NRR/ DST /GIB U5! [ NOTE 3(a)] I 6/30/85 A-32 Missile Effects Pittman NRR/DE/MTEB A-37, A-38, 11/30/83 NA B-68 A-33 NEPA Review of Accident Risks -
NRR/D5I/AEB 11/30/83 E(NOTE 3) NA A-34 Instruments for Monitoring Radiation and Process V'Molen NRR/051/IC58 II.F.3 11/30/83 NA Variables During Accidents A-35 Adequacy of Offsite Power Systems Enrit NRR/DSI/PSB 11/30/83 NOTE 3(a)
A-36 Control of Heavy Loads Near Spent Fuel -
NRR/DSI/GIB 6/30/85 USI [ NOTE 3(a)] 1 C-10, C-15
$ A-37 A-38 Turbine Missiles Pittman NRR/DE/MTEB DROP 11/30/83 NA Tornado Missiles Sege NRR/DSI/ASB LOW 11/30/83 NA A-39 Determination of Safety Relief Valve Pool Dynamic -
NRR/ DST /GIB 6/30/85 USI [ NOTE 3(a)] 1 Loads and Temperature Limits A-40 Seismic Design Criteria - Short Term Program -
NRR/ DST /GIB USI . 11/30/83 A-41 Long Tere Seismic Program Colmar NRR/DE/MEB 12/31/84 NOTE 3(b) 1 NA A-42 Pipe Cracks in Boiling Water Reactors -
NRR/ DST /GIB USI [ NOTE 3(a)] I 6/30/85 8-05 A-43 Containment Emergency Sump Performance -
NRR/ DST /CIB USI 11/30/83 A-44 Station Blackout -
NRR/ DST /CIB USI 11/30/83 A-45 Shutdown Decay Heat Removal Requirements -
NRR/ DST /GIB USI 11/30/83 A-46 Seismic Qualification of Equipment in Operating Plants -
NRR/ DST /GIB USI 11/30/83 A-47 Safety Implications of Control Systems -
NRR/ DST /GIB USI 11/30/83 A-48 Hydrogen Control Measures and Effects of Hydrogen Burns -
NRR/ DST /GIB USI 11/30/83 on Safety Equipment A-49 Pressurized Thermal Shock -
NRR/ DST /GIB USI 11/30/83 8-1 Environmental Technical Specifications -
NRR/DE/EHEB 11/30/83 E (NOTE 3) NA B-2 Forecasting Electricity Demand -
NRR E (NOTE 3) 11/30/83 NA B-3 Event Categorization -
NRR/DSI/R5B LI (DROP) 11/30/83 NA y B-4 ECCS Reliability Enrit NRR/DSI/RSB II.E.3.2 11/30/83 NA m B-5 Ductility of Two Way Slabs and Shells and Buckling Thatcher Q Behavior of Steel Containments NRR/DE/5GEB MEDIUM 11/30/83 g B-6 Loads, Load Combinations, Stress Limits ' <
e Pittman NRR/DE/MEB HIGH 11/30/83 -**
B-7 Q
w 8-8 Secondary Accident Consequence Modeling Locking Out of ECCS Power Operated Valves Riggs NRR/DSI/AEB LI (DROP) 11/30/83 NA $
W NRR/DSI/RSB DROP 11/30/83 NA O B-9 Electrical Cable Penetrations of Containment Emrit NRR/DSI/PSB 11/30/83 3 NOTE 3(b) NA B-10 Behavior of BWR Mark III Containments V'Molen NRR/DSI/CSB NOTE 3(a) 1 12/31/84 NA m B-11 Subcompartment Standard Problems NRP/DSI/CSB 11/30/83 LI (NOTE 5) NA
o
(
w TABLE II (Continued) o Action Lead Lead Office / Safety latest om Plan Item / SPEB Division / Priority latest Issuance MPA Issue No. Title Engineer ' Branch Ranking Revision Date No.
B-12 Containment Cooling Requirements (Non-LDCA) Emrit NRR/D5I/CSB NOTE 3(a) 11/30/83 B-13 Marviken Test Data Evaluation -
NRR/051/C58 LI (NOTE 5) 11/30/83 NA B-14 Study of Hydrogen Mixing Capability in Containment Emrit NRR/ DST /CIB A-48 11/30/83 NA Post-LOCA B-15 CONTEMPT Computer Code Maintenance -
NRR/DSI/CSB LI (DROP) 11/30/83 NA B-16 Protection Against Postulated Piping failures in Fluid Emrit NRR/DE/MEB A-18 11/30/83 NA Systems Outside Containment B-17 Criteria for Safety-Related Operator Actions Milstead NRR/DHF5/LQB HF 01. 4. 3 1 12/31/85 NA B-1B Vortex Suppression Requirements for Containment Sumps Enrit NRR/ DST /GIB A-43 11/30/83 NA B-19 Thermal-Hydraulic Stability Colmar NRR/DSI/CPB NOTE 3(b) 6/30/85 NA B-20 Standard Problem Analysis -
RES/DAE/AMBR LI (NOTE 5) 11/30/83 B-21 Core Physics -
NRR/DSI/CPB LI (DROP) 11/30/83 NA B-22 LWR Fuel V'Molen NRR/D51/CPB NOTE 4 11/30/83 B-23 LMFBR Fuel -
NRR/051/CPB LI (DROP) 11/30/83 NA B-24 Seismic Qualification of Electrical and Mechanical Emrit NRR A-46 11/30/83 NA Components B-25 Piping Benchmark Problems -
NRR/DE/MEB LI (NOTE 5) 11/30/83 8-26 Structural Integrity of Containment Penetrations Riggs NRR/DE/MTEB NOTE 3(b) 1 12/31/84 NA B-27 Implementation and Use of Subsection NF -
NRR/DE/MEB L1 (NOTE 5) 11/30/83 g B-28 Radionuclide/ Sediment Transport Program -
NRR/DE/EHEB E (NOTE 3) 11/30/83 NA B-29 Effectiveness of Ultimate Heat Sinks Pittman NRR/DE/EHEB NOTE 4 11/30/83 B-30 Design Basis Floods and Probability -
NRR/DE/EHEB L1 (NOTE 5) 11/30/83 B-31 Dam failure Model Milstead NRR/DE/SCEB NOTE 4 11/30/83 B-32 Ice Effects on Safety Related Water Supplies Milstead NRR/DE/EHEB NOTE 4 11/30/83 B-33 Dose Assessment Methodology -
NRR/DSI/RAB LI (NOTE 3) 11/30/83 NA B-34 Occupational Radiation Exposure Reduction Emrit NRR/DSI/RAB Ill.D.3.1 11/30/83 NA B-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 from 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 Normal Ventilation Systems B-37 Chemical Discherges to Receiving Waters -
NRR/DE/EHEB E (NOTE 5) 11/30/83 B-38 Reconnaissance Level Investigations -
NRR/DE/EHEB E (DROP) 11/30/83 NA B-39 Transmission Lines -
NRR/DE/EHEB E (DROP) 11/30/83 NA B-40 Effects of Power Plant Entrainment on Plankton -
NRR/DE/EHEB E (DROP) 11/30/83 NA 3E B-41 Impacts on Fisheries -
NRR/DE/E HEB E LOROP) 11/30/83 NA E
m B-42 B-43 Socioeconomic Environmental Impacts -
NRR/DE/SAB E (NOTE 3) 11/30/83 NA 33 Value of Aerial Photographs for Site Evaluation -
NRR/DE/EHEB E (NOTE 5) 11/30/83 @
7o B-44 Forecasts of Generating Costs of Coal and Nuclear Plants NRR/DE/SAB E (NOTE 3) 11/30/83 NA -*-
w B-45 B-46 Naed f or Power - Energy Conservation Cost of Alternatives in Environmental Design NRR/DE/5AB NRR/DE/SAB E (B-2) 11/30/83 11/30/B3 NA NA o
3 E (DROP)
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g TABLE II (Continued) k cn Action Lead lead Office / Safety Latest m Plan Item / SPEB Division / Priority latest I s suanc e MPA Issue No. Title Engineer Branch Ranking Revision Date No.
B-47 Inservice Inspection of Supports-Classes 1, 2, 3, and Colmar NRR/DE/MIEB DROP 11/30/83 NA HC Components B-48 BWR CRD Mechanical Failure (Collet Housing) Emrit NRR/DE/Mi[B NOTE 3(b) 11/30/83 8-49 Inservice Inspection Criteria and Corrosion Prevention -
NRR LI (NOIE 5) 11/30/83 Criteria for Containments B-50 Post-Operating Basis Earthquake Inspection Colmar NRR/DE/SGEB RI (LOW) 1 06/30/85 NA B-51 Assessment of Inelastic Analysis Techniques for Emrit NRR/DE/MEB A-40 11/30/83 NA Equipment and Components B-52 Fuel Assembly Seismic and LOCA Responses Emrit NRR/ DST /GIB A-2 11/30/83 NA -
B-53 Load Break Switch Sege NRR/DSI/PSB RI (NOTE 3) 11/30/83 B-54 Ice Condenser Containments Milstead NRR/DSI/CSB NOTE 3(b) 1 12/31/84 NA B-55 Improved Reliability of Target Rock Safety Relief V'Molen NRR/DE/MEB MEDIUM 11/30/83 Valves B-56 Diesel Reliability Milstead NRR/DSI/PSB HIGH 11/30/83 D-19 B-57 Station Blackout' Enrit NRR/ DST /GIB A-44 11/30/83 B-56 Passive Mechanical Failures Colmar NRR/DE/EQB NOTE 3(b) 1 12/31/85 NA B-59 (N-1) Loop Operation in BWRs and PWRs Colmar NRR/DSI/R$8 RI (NOTE 3) 1 6/30/85 E-04,E-05 B-60 Loose Parts Monitoring System Enrit NRR/DSI/CPB NOTE 3(b) 1 12/31/84 NA B-61 Allowable ECCS Equipment Outage Periods Pittman NRR/ DST /RRAB MEDIUM 11/30/83
$ B-62 Reexamination of Technical Bases for Establishing SLs, LSSSs, and Reactor Protection System Trip functions NRR/DSI/CPB LI (DROP) 11/30/83 NA B-63 Isolation of Low Pressure Systems Connected to the Emrit NRR/DE/MEB NOTE 3(a) 11/30/83 Reactor Coolant Pressure Boundary B-64 Decommissioning of Reactors Colmar. NRR/DE/CHEB NOTE 2 11/30/83 B-65 Iodine Spiking Milstead NRR/DSI/AEB CROP 2 12/31/84 NA B-66 Control Room Infiltration Measurements Matthews NRR/DSI/AEB NOTE 3(a) 11/30/83 B-67 Effluent and Process Monitoring Instrumentation Colmar NRR/DSI/METB III.D.2.1 11/30/83 NA B-68 Pump Overspeed During LOCA Riani NRR/DSI/ASB DROP 11/30/83 NA B-69 ECCS Leakage Ex-Containment Riaqi NRR/DSI/METB 111. D.1.1 11/30/83 NA B-70 Power Grid Frequency Degradation and Effect on Primary . Enrit NRR/DSI/PSB NOTE 3(a) 11/30/83 Coolant Pumps B-71 Incident Response Riani NRR III.A.3.1 11/30/83 NA B-72 Health Effects and Life Shortening from Uranium and -
NRR/DSI/RA8 LI (NOTE 5) 11/30/83 NA Coal Fuel Cycles B-73 Monitoring for Excessive Vibration Inside the Reactor Thatcher NRR/DE/MEB C-12 11/30/83 NA Pressure Vessel C-1 Assurance of Continuous Long Term Capability of Hermetic Milstead NRR/DE/EQB NOTE 3(a) 11/30/83 y Seals on Instrumentation and Electrical Equipment m C-2 Study of Containment Depressurization by Inadvertent Emrit NRR/DSI/CSB NOTE 3(b) 11/30/83 NA m y Spray Operation to Determine Adequacy of Containment e O
s External Design Pressure 1 C-3 Insulation Usage Within Containment Enrit NRR/ DST /GIB A-43 11/30/83 NA et 8 C-4 Statistical Methods for ECCS Analysis Riggs- NRR/DSR0/SPEB RI (NOTE 3) 1 06/30/86 NA g (d C-5 Decay Heat Update Riggs NRR/DSRO/SPEB RI (NOTE 3) 1 06/30/86 NA 3 un
o c)
D o
TABLE II (Continued)
N 00 Action Lead lead Office / Safety latest
- Plan Item / SPEB Division / Priority latest Issuance MPA Issue No. Title Engineer Branch Rcnking Revision Date No.
C-6 LOCA Heat Sources Riggs NRR/DSR0/SPEB RI (NOTE 3) 1 06/30/86 NA C-7 PWR System Piping Emrit NRR/DE/MTEB NOTE 3(b) 11/30/83 NA C-8 Main Steam Line Leakage Control Systems Milstead NRR/DSI/ASB HIGH 11/30/83 C-9 RHR Heat Exchanger Tube Failures V'Molen NRR/DSI/RSB DROP 11/30/83 NA C-10 Effective Operation of Containment Sprays in a LOCA Emrit NRR/DSI/AEB NOTE 3(a) 11/30/83 NA C-Il Assessment of Failure and Reliability of Pumps and Emrit NRR/DE/MEB NOTE 3(b) 12/31/85 NA valves C-12 Primary System Vibration Assessment Thatcher NRR/DE/MEB NOTE 3(b) 11/30/83 NA C-13 Non-Random Failures Enrit NRR/ DST /GIB A-17 11/30/83 NA C-14 Storm Surge Model for Coastal Sites Emrit NRR/DE/EHEB NOTE 4 11/30/83 C-15 NUREG Report for Liquids Tank Failure Analysis -
NRR/DE/EHEB LI (DROP) II/3C/83 NA C-16 Assessment of Agricultural Land in Relation to Power -
NRR/DE/EHEB E (DROP) 11/30/83 NA Plant Siting and Cooling System Selection C-17 Interim Acceptance Criteria for Solidification Agents Emrit NRR/DSI/MtTB NOTE 3(a) 11/30/83 NA for Radioactive Solid Wastes 0-1 Advisability of a Seismic Scram Thatcher RES/DET/MSEB LOW 11/30/83 NA D-2 Emergency Core Cooling System Capability for Future Emrit hRR/DSI/R$B NOTE 4 11/30/83 Plants D-3 b ntrol Rod Drop Accident Enrit NfR/DSI/CPB NOTE 3(b) 11/30/83 NA 3
ro NEW GENERIC ISSUES
- 1. Failures in Air-Mor.itoring, Air-Cleaning, and Emrit NRR/DSI/METB DROP 11/30/83 NA Ventilating Systems
- 2. Failure of Protective Devices on Essential Equipment Colmar NRR/DSI/ICSB NOTE 4 11/30/83 NA
- 3. Set Point Drift in Instrumentation Emrit NRR/DSR0/RSIB NOTE 3(b) 1 06/30/86 NA
- 4. End-of-Life and Maintenance Criteria Thatcher NRR/DE/EQB NOTE 3(b) 11/30/83 NA
- 5. Design Check and Audit of Balance-of-Plant Equipment Pittman NRR/DSI/ASB I.F.1 11/30/83 NA
- 6. Separation of Control Rod from Its Drive and BWR High V'Molen NRR/DSI/CPB NOTE 3(b) 11/30/83 NA Rod Worth Events
- 7. Failures Due to Flow-Induced Vibrations V'Molen NRR/DSI/R$8 DROP 11/30/83 NA
- 8. Inadvertent Actuation of Safety Injection in PWRs Colmar NRR/DSI/RSB I.C.1 11/30/83 NA
- 9. Reevaluation of Reactor Coolant Pump Trip Criteria Emrit NRR/DSI/RSB II.K.3(S) 11/30/83 NA
- 10. Surveillance and Maintenance of TIP Isolation Valves Riggs NRR/DSI/ICSB DROP 11/30/83 NA 2 and Squib Charges Turbine Disc Cracking Pittman NRR/DE/MTEB A-37 11/30/83 NA rrt 11.
- 12. BWR Jet Pump Integrity Sege NRR/DE/MTEB NOTE 3(b) 1 12/31/84 NA [
o 13. Small Break LOCA from Extended Overheating of Riani MLB NRR/DSI/RSB DROP 11/30/83 NA 7
$ Pressurizer Heaters -*.
(> 14. PWR Pipe Cracks Emrit NRR/DE/MTEB No:E 3(b) 1 12/31/85 NA @
- 15. Radiation Effects on Reactor Vessel Supports Emrit NRR/DE/MIEB LOW 11/30/83 NA tsn 9 O ,
O
. _ = ._. _. .__ . . _ . _ . . _ _ _ . . m k
j J J o
m N . TABLE II (Continued) w O
h m
Action Plan Item /
Lead Lead Office / Safety latest SPEB . Division / Priority Latest issuance MPA issue No. Title Engineer Branch , Ranking Revision Date No.
- 16. BWR Main Steam isolation Valve Leakage Control Systems Milstead NRR/D5I/ASB C-8 11/30/83 NA
- 17, Loss of Offsite Power Subsequent to LOCA Colmar NRR/051/P58, DROP 11/30/83 NA i
ICSB
- 18. Steam Line Break with Consequential Small LOCA Riggs NRR/D5I/RSB 1.C.1 11/30/83 NA
- 19. Safety Implications of Nonsafety Instrument and Control Sege NRR/ DST /GIB A-47 11/30/83 NA Power Supply Bus
- 20. Effects of Electromagnetic Pulse on Nuclear Power Thatcher NRR/051/ICSB NOTE 3(b) 1 06/30/84 NA Plants t
- 21. Vibration Qualification of Equipment Riggs NRR/DE/EQB DROP 1 06/30/86 NA
- 22. Inadvertent Boron Dilution Events V'Molen NRR/DSI/RSB NOTE 3(b) 1 12/31/84 NA
- 23. Reactor Coolant Pump Seal Failures Riggs NRR/DSI/ASB HIGH 11/30/83
- 24. Automatic Emergency Core Cooling System Switch to V'Molen NRR/OSI/RSB NOTE 4 11/30/83 Recirculation
! 25. Automatic Air Header Dump on BWR Scram System Milstead NRR/DSI/R58 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/R5B B-17 11/30/83 NA
- 28. Pressurized Thermal Shock Enrit NRR/ DST /GIB A-49 11/30/83 NA
- 29. Bolting Degradation or Failure in Nuclear Power Plants V'Molen NRR/DE/MTEB HIGH 11/30/83
$ 30. Potential Generator Missiles - Generator Rotor Retaining Rings Pittman NRR/DE/MEB DROP 1 12/31/85 NA
- 31. Natural Circulation Cooldown Riggs NRR/DSI/R$B I . C.1 11/30/83 NA i
- 32. Flow Blockage in Essential Equipment Caused by Corbicula Emrit NRR/D51/A58 51 11/30/83 NA l 33. Correcting Atmospheric Dump Valve Opening Upon Loss of Pittman NRR/DSI/IC58 A-47 11/30/83 NA Integrated Control System Power
, 34. RCS Leak Riggs NRR/DHFS/PSRB DROP 1 06/30/84 NA
] 35. Degradation of Internal Appurtenances in LWRs. V'Molen NRR/DSI/CPB, LOW l 06/30/85 NA a
R58
- 36. Loss of Service Water Colmar NRR/DS!/ASB, NOTE 3(b) 2 06/30/86 NA AEB, RSB
! 37. Steam Generator Overfill and Combined Primary and Colmar NRR/ DST /GIB, A-47, 1 06/30/85 NA
! Secondary Blowdown NRR/051/R5B I.C.1 1 38. Potential Recirculation System Failure as a Consequence Milstead NRR NOTE 4 11/30/83 of Injection of Containment Paint Flakes or Other Fine Debris
- 39. Potential for Unacceptable Interaction Between the CR0 Pittman NRR/DSI/ASB 25 11/30/83 NA 2 System and Non-Essential Control Air System
- 5 rn
- 40. Safety Concerns Associated with Pipe Breaks in the BWR Colmar Scram System NRR/DSI/ASB NOTE 3(a) 1 06/30/84 B-65 :o
] @ 41. BWR Scram Discharge Volume Systems V'Molen NRR/051/R58 h0TE 3(a) 11/30/83 8 58 1-g 42. Combination Primary / Secondary System LOCA Riggs NRR/DSI/RSB I.C.1 'l - 06/30/85 NA
{
w 43. Contamination of Instrument Air Lines Milstead NRR/051/ASB DROP 11/30/83 NA o W 44. Failure of Saltwater Cooling System Milstead NRR/DSI/ASB 43 11/30/83 NA 3 U1
o
(
w TABLE II (Continued) o y Action Lead Lead Office / Safety Latest am Plan Item / SPEB Division / Priority latest Issuance MPA Issue No. Title Engineer Branch Ranking Revision Date No.
- 45. Inoperability of Instrumentation Due to Extreme Cold Milstead NRR/DSI/ICSB NOTE 3(a) 1 06/30/84 Weather 46, toss of 125 Volt DC Bus Sege NRR/DSI/PSB 76 11/30/83 NA
- 47. Loss of Off-site Power lhatcher NRR/051/R$B, NOTE 3(b) 11/30/83 ASS
- 48. LCO for Class IE Vital Instrument Buses in Operating Sege NRR/DSI/PSB NOTE 2 11/30/83 Reactors
- 49. Interlocks and LCOs for Redundant Class IE Tie Breakers Sege NRR/DSI/PSB MEDIUM 1 12/31/84 Sd. Reactor Vessel Level Instrumentation in BWRs Thatcher NRR/DSI/R5B, NOTE 3(b) 1 12/31/84 NA ICSB
- 51. Proposed Requirements for Improving the Reliab.ility of Emrit NRR/DSI/ASB MEDIUM 11/30/83 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 Blockage Incident V'Molen NRR/DSI/CPB, OROP 1 12/31/84 NA in a BWR R5B
- 54. Valve Operator-Related Events Occurring During 1978, Colmar NRR/DE/MEB II.E.6.1 1 06/30/85 NA 1979, and 1980
- 55. Failure of Class IE Safety-Related Switchgear Circuit Emrit NRR/DSI/PSB DROP 1 12/31/85 NA Breakers to Close on Demand g 56. Abnormal Transient Operating Guidelines as Applied to a Steam Generator Overfill Event Colmar NRR/DHFS/HFEB A-47, I.D.1 11/30/83 NA
- 57. Effects of Fire Protection System Actuation Milstead NRR NOTE 4 11/30/83 on Safety-Related Equipment
- 58. Inadvertent Containment Flooding Sege NRR/D5I/ASB, DROP 11/30/83 CSB
- 59. Technical Specification Requirements for Plant Shutdown Emrit 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 Wetwell Airspace of Mark I Milstead NRR/DSI/CSB MEDIUM 1 12/31/85 and II Containments
- 62. Reactor Systems Bolting Applications V'Molen NRR NOTE 4 11/30/83
- 63. Use of Equipment Not Classified as Essential to Safety V'Molen NRR NOTE 4 11/30/83 in BWR Transient. Analysis
- 64. Identification of Protection System Instrument sensing Thatcher NRR/DS!/ICSB NOTE 3(b) 11/30/83 Lines
- 65. Probability of Core-Melt Due to Component Cooling Water V'Molen NRR/DSI/ASB HIGH 11/30/83 2 System Failures E
rn 66.
67.
Steam Generator Re wirements Steam Generator Staff Actions Riggs NRR/DL/0RAB NOTE 2 1 06/30/85 m
n>
Q 67.2.1 67.3.1 Integrity of Steam Generator Tube Sleeves Riggs NRR/DE/MEB R1 (NOTE 5)
A-47, 1 06/30/85 06/30/85 NA 1 o Steam Generator Overfil! Riggs NRR/ DST /GIB 1 NA un 8 NRR/DSI/R5B I.C. I g w 67.3.2 Pressurized Thermal Shock Riggs NRR/D5T/GIB A-49 1 06/30/85 NA :s 67.3.3 Improved Accident Monitoring Riggs NRR/DSI/lCSB NOTE 3(a) 1 06/30/85 A-17 g O O O
A g
(%
t
\
'v'
/
o
(
w TABLE II'(Continued) o y Action Plan item /
Lead Lead Office / Safety latest Iss uance MPA m SPEB Division / Priority Latest lssue No. Title Engineer Branch Ranking Revision Date No.
67.3.4 Reactor Vessel Inventory Measurement Riggs NRR/DSI/CPB II.f.2 1 6/30/85 NA 67.4.1 RCP Trip Riggs NRR/D51/R58 II.K.3(5) 1 6/30/85 NA 67.4.2 Control Room Design Review Riggs NRR/DHFS/HfEB I.D.1 1 6/30/85 NA 67.4.3 Emergency Operating Procedures Riggs NRC/DHF5/PSRB I.C.1 1 6/30/85 NA 67.5.1 Reassessment of SGTR Design Basis Riggs NRC/DSI/AEB L1 (NOTE 5) 1 6/30/85 NA 67.5.2 Reevaluation of SGTR Design Basis Riggs NRR/DSI/RSB L1 (NOTE 5) 1 6/30/85 NA 67.5.3 Secondary System Isolation Riggs NRR/DSI/RSB DROP 1 6/30/85 NA 67.6.0 Organizational Responses Riggs ole /DEPER/IRDB Ill.A.3 1 6/30/85 NA 67.7.0 Improved Eddy Current Tests Riggs NRR/DE/MTEB MEDIUM 1 '6/30/85 67.8.0 Denting Criteria Riggs NRR/DE/MfEB RI (NOTE 5) 1 6/30/85 NA 67.9.0 Reactor Coolant System Pressure Control Riggs NRR/DSI/GIB A-45, 1 6/30/85' NA NRR/DSI/RSB 1.C.1 67.10.0 Supplement Tube-Inspections Riggs NRR/DL/0RAB LI (NOTE 5) 1 6/30/85 NA
- 68. Postulated Loss of Auxiliary Feedwater System Resulting 'Pittman NRR/DSI/ASB HIGH 1 6/30/84 from Turbine-Driven Auxiliary Feedwater Pump Steam Supply Line Rupture
- 69. Make-up Nozzle Cracking in B&W Plants Colmar NRR/DE/MEB, NOTE 3(b) 1 12/31/64 (later)
HTEB
- 70. PORV and Block Valve Reliability Riggs NRR/DSI/R58 MEDIUM 1 6/30/84 A 71. Failure of Resin Demineralizer Systems and Their Emrit NRR NOTE 4 11/30/83
- Effects on Nuclear Power Plant Safety
- 72. Control Rod Drive Guide Tube Support Pin Failures V'Molen NRR NOTE 4 11/30/83
- 74. Reactor Coolant Activity Limits for Operating Reactors Milstead NRR/DSI/AEB DROP 1 06/30/86 NA
- 75. Generic Implications of ATWS Events at the Salem Thatcher NRR/DSI NOTE 1 11/30/83 B-76,8-77 Nuclear Plant B-78,8-79 B-60,B-81 B-82,B-85 B-86,8-87 B-88,B-89 B-90,B-91 B-92,B-93
- 76. Instrumentation and Control Power Interactions Colmar NRR NOTE 4 11/30/83
- 77. Flooding of Safety Equipment Compartments by Back-flow Colmar NRR/DSI/ASB HIGH 11/30/83 Through Floor Drains
- 78. Monitoring of Fatigue Transient Limits for Reactor, Riggs NRR NOTE 4 11/30/83 Coolant System z 79. Unanalyzed Reactor Vessel Thermal Stress During Colnar NRR/DE/MEB, MEDIUM 1 12/31/84
% Natural Convection Cooldown NRR/DSI/RSB ,
i rri 80. Pipe Break Effects on Control Rod Drive Hydraulic Lines V'Holen NRR/DSI/R58, LOW 11/30/83 NA e
@ in the Drywells of BWR Mark I and II Containments ASB, $
m o CP8 j $ 81. Impact of Locked Doors and Barriers on Plant Personnel Colmar NRR/DHPS/PSRB DROP 1 12/31/84 NA y
.; w and Safety -s
- 82. Beyond Design Basis Accidents in Spent Fuel Pools V'Molen NRR/DSI/AEB MEDIUM 11/30/83 ,
- 83. Control Room Habitability Matthews NRR NOTE 4 11/30/83
- 85. Reliability of Vacuum Breakers Connected to Steam Milstead NRR/DSI/C58 DROP 1 12/31/85 NA Discharge Lines Inside BWR Containments
o
(
w FABLE II (Continued) o y Action Lead Lead Office / Safety latest cn Plan Item / SPEB Division / Priority Latest Issuance MPA Issue No. Title Engineer Branch Ranking Revision Date No.
- 86. Long Range Plan for Dealing with Stress Corrosion Emrit NRR/DE/MIEB NOTE 2 12/31/84 B-84 Cracking in BWR Piping
- 87. Failure of HPCI Steam Line Without Isolation Pittman hRR/051/R58 88.
HIGH 12/31/85 Earthquakes and Emergency Planning Emrit NRR NOTE 4 (later)
- 89. Stiff Pipe Clamps Riggs
- 90. Technical Specifications for Anticipatory Trips NRR NOTE 4 (later)
V'Molen NRR/DSI/RSB, LOW 12/31/84 NA ICSB .
- 91. Main Crankshaft Failures in Transamerica DeLaval Emrit NRR/DL NOTE 1 12/31/85 Emergency Diesel Generators
- 93. Steam Binding of Auxiliary Feedwater Purpps Pittman NRR/DSI/ASB 94.
HIGH 12/31/84 Additional Low Temperature Overpressure Protection Pittman NRR/05!/RSB HIGH 13/31/85 Issues for Light Water Reactors
- 95. Loss of Effective Volume for Containment Recirculation Milstead Spray NRR NOTE 4 (later)
- 96. RHR Suction Valve Testing Milstead
V*Molen NRR/DSI/RAB III.D.3.1 06/30/85 NA
- 98. CRD Accumulator Check Valve Leakage Pittman NRR/DSI/ASB A 99.
DROP 06/30/85 NA
- RCS/RHR Suction Line Valve Interlock on PWRs Pittman NRR/051/R58 HIGH 1 06/30/86 100. 015G Level Riggs 101. Break Plus Single Failure in BWR Water Level NRR NOTE 4 (later)
V'Molen NRR/DSI/RSB HIGH 06/30/85 Instrumentation 102. Human Error in Events Involving Wrong Unit or Wrong Emrit NRR/DHF5/LQB HF02 06/30/85 NA Train 103. Design for Probable Maximum Precipitation Emrit NRR/DE/EHEB 104.
NOTE 1 12/31/85 Reduction of Boron Dilution Requirements V'Molen 105. Interfacing Systems LOCA at BWRs NRR NOTE 4 (later)
Milstead NRR/051/R58 HIGH 06/30/85 106. Piping and Use of Highly Combustible Gases in Vital Colmar Areas NRR NOTE 4 (later) 10/. Generic Implications of Main Transformer Failures Colmar 108. BWR Suppression Pool Temperature Limits NRR NOTE 4 (later)
Colmar NRR/DSI/CSB RI (LOW) 06/30/85 NA 109. Reactor vessel Closure Failure V'Molen NRR NOTE 4 (later) 110. Equipment Protective Devices on Engineered Safety Milstead Features NRR NOTE 4 (later) 111. Stress Corrosion Cracking of Pressure Boundary Riggs NRR/DE/MTEB LI (NOTE 5) 12/31/85 NA Ferritic Steels in Selected Environments 2 112. Westinghouse RPS Surveillance frequencies and Pittman NRR/DSI/ICSB RI (NOTE 3) 12/31/85 NA E
m 113.
Out-of-service. Times m
Dynamic Qualification Testing of Large Bore Riggs NRR NOTE 4 (later) fD 9
o 114.
Hydraulic Snubbers Seismic-Induced Relay Chatter Riggs NRR/05R0/SPEB A-46 06/30/86 NA to 8
w 115. Reliability of Westinghouse Solid State Protection System Milstead NRR NOTE 4 (later) 8
- 3 116. Accident Management Pittman NRR/DHf5 NOTE 4 (later) m O O O
. ._ _ _ _ _ _ _ . _. m _ _
N .
% %) ,
o cn g TA8LE II (Continued)
R co Action Lead Lead Office / Safety latest Ch Plan item / SPEB Division / Priority latest Issuance MPA Issue No. Title Engineer Branch ' Ranking Revision Date- No.
117. Allowable Outage Times for Diverse Simultaneous NRR NOTE 4 (later)
Equipment Outages 118. Tendon Anchorage Failure Milstead NRR NOTE 4 (later) 119.. Piping Review Committee Recommendations - - -
119.1 Piping Rupture Requirements and Decoupling of Riggs NRR/DE RI (NOTE 5) 12/31/85 NA Seismic and LOCA Loads 119.2 Piping Damping Values Riggs NRR/DE RI (NOTE 5) 12/31/85 NA 119.3 Decoupling the OBE from the SSE Riggs NRR/DE RI (NOTE 5) 12/31/85 NA 119.4 BWR Piping Materials Riggs NRR/DE RI (NOTE 5) 12/31/85 NA 119.5 . Leak Detection Requirements Riggs a NRR/DE RI (NOTE 5) 12/31/85 NA 120. On-Line Testability of Protection Systems Milstead NRR NOTE 4 (later) 121. Hydrogen Control for Large, Dry PWR Containments Emrit NRR HIGH 12/31/85 4 122. Davis-Besse loss of All Feedwater Event of June 9 - - -
1985 - Short-Term Actions 122.1 . Potential Inability to Remove Reactor Decay Heat - - -
122.1.a ~ Failure of Isolation Valves in Closed Position V'Molen NRR/DSR0/RSIB HIGH 06/30/86 l 122 1.b Recovery of Auxiliary Feedwater V'Molen NRR/DSR0/RSIB MEDIUM 06/30/86 122.1.c. Interruption of Auxiliary Feedwater Flow V'Molen NRR/DSR0/RSIB HIGH 06/30/86 D -122.2 Initiating Feed-and-Bleed . V'Molen NRR/DHFT HIGH 06/30/86 4 122.3 Physical Security System Constraints V'Molen NRR LOW 06/30/86 123. Deficiencies in the Regulations Governing DBA and Rowsome NRR NOTE 4 (later)
Single-Failure Criteria Suggested by the Davis-Besse Event of June 9, 1985 124. Auxiliary feedwater System Reliability Entit NRR/DSRO/RSIB NOTE 1 06/30/86 125. Davis-Besse Loss of All Feedwater Event of - - -
June 9, 1985 - Long-Term Actions 125.I.1 Availability of the STA V'Molen NRR/ NOTE 4 (later) 125.I.2 PORV Reliability - - - -
125.I.2.a Need for a Test Program to Establish Reliability of V'Molen NRR/DSRO/SPEB 70 06/30/86 NA the PORV 125.I.2.b Need for PORV Surveillance Tests to Confirm V*Molen NRR/DSR0/SPEB 70 06/30/86 NA Operational Readiness 2 125.I.2.c c Need for Additional Protection Against PORV Failure V'Molen NRR/DSR0/SPEB DROP 06/30/86 NA
- :o 125.I.2.d Capability of the PORV to Support feed-and-Bleed V'Molen NRR/DSR0/SPE8 A-45 06/30/86 NA :g, Q 125.I.3 SPDS Availability Milstead NRR/. NOTE 4 (later) a 125.I.4 Plant-Specific Simulator Riggs NRR/ NOTE 4 (later)
' .g
@ 125.1.5 Safety Systams Tested in All Conditions Required by V'Molen NRR/ NOTE 4 (later) w Design Basis Analysis {
o W 125.I.6 Valve Torque Limit and Bypass Switch Settings V'Moleni
' NRR/ N01E 4 (later). :s 125.I.7 Operator Training Adequacy - - -
m
o
(
w TABLE II (Continued) o y Action lead Lead Office / Safety latest cn Plan item / SPEB Division / Priority latest Issuance MPA Issue No. Title Engineer Branch Ranking Revision Date No.
125.I.7.a Recover f ailed Equipment V'Holen NRR/ NOTE 4 (later) 125.1.7.b Realistic Hands-On Training V'Mclen NRR/ NOTE 4 (later) 125.I.8 Procedures and Staffing for Reporting to NRC Emergency V*Molen NRR/ NOTE 4 (later)
Response Center 125.II.1 AFW System Evaluation - - -
125.II.1.a Two-Train AfW Reliability V'Molen NRR/ NOTE 4 (later) 125.II.I.b Review Existing AfW Systems for Single failure V'Molen NRR/ NOTE 4 (later) 125.II.I.c NUREG-0737 Reliability Improvements V'Molen NRR/ NOTE 4 (later) 125.II.l.d AfW/ Steam and feedwater Rupture Control System /ICS V'Molen NRR/ NOTE 4 (later)
Interactions in B&W Plants 125.11.2 Adequacy of Existing Maintenance Requirements for Riggs NRR/ N01E 4 (later)
Safety-Related Systems 125.II.3 Review Steam / feed Line Break Mitigation Systems for V'Molen NRR/ NOTE 4 (later)
Single failure 125.11.4 OT5G Dryout and Reflood Effects Riggs NRR/ NOTE 4 (later) 125.11.5 1hermal-Hydraulic Effects of Loss and Restoration Riggs NRR/ N01E 4 (later) of feedwater on Primary System Components 125.!!.6 Reexamine PRA Estimates of Core Damage Risk from V'Molen NRR/ NOTE 4 (later)
Loss of All feedwater g 125.11.7 Reevaluate Provisions to Automatically Isolate feedwater from Steam Generator During Line Break V'Molen NRR/ NOTE 4 (later) 125.11.8 Reassess Criteria for feed-and-Bleed Initiation V'Molen NRR/ NOTE 4 (later) 125.11.9 Enhance Feed-and-Bleed Capability .V'Molen NRR/ NOTE 4 (later) 125.11.10 Hierarchy of Impromptu Operator Actions Riggs NRR/ NOTE 4 (later) 125.11.11 Recovery of Main feedwater as Alternative to AfW V'Molen NRR/ N01E 4 (later) 125.11.12 Adequacy of Training Regarding PORV Operation Riggs NRR/ NOTE 4 (later) 125.11.13 Operator Job Aids Pittman NRR/ N01E 4 (later) 125.II.14 Remote Operation of Equipment Which Must Now Be V'Molen NRR/ NOTE 4 (later)
Operated locally 126. Reliability of PWR Main Steam Safety Valves Riggs NRR/. NOTE 4 (later) 121. Testing and Maintenance of Manual Valves in Safety- Pittman NRR/ NOTE 4 (later)
Related Systems 128. Electrical Power Reliability Emrit NRR/ NOTE 4 (later) 129. Valve Interlocks to Prevent Vessel Drainage During Milstead NRR/ N01E 4 (later)
Shutdown Coolind 130. Essential Service Water Pump f ailures at Multiplant Riggs NRR/ NOTE 4 (later)
Sites 2
E rn HUMAN FACTORS 1550E5 m (D
O <
Hf01 7 f
m -
HUMAN FACIORS PROGRAM PLAN (HFPP) w o w HF 01.1. 0 Staffing and Qualifications - - -
- 3 Hf01.1.1 Policy Statement on [ngineering Expertise on Shift Pittman NRR/DHf5/LQB HIGH 12/31/84 m and Evaluate Effectiveness of Policy Statement O O O
(. , f~
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O Ch FAB 1E Il J ontinued) o ___ _ _ . . _ _
CD Action lead lead Of f ice-/ Sat.5ty latest CD Plan item / SPEB Division / Prsurity latest I ssaiin . MPA Issue No. Iitle Engineer Bran (h N mL i nig Revision Date No.
HF 01.1. 2 Revise and Evaluate Changes to Regulatory Guide 1.8 Pittman NRR/DHF 5/LQB HIGH ]?/31/84 Nf 01.1. 3 Develop a Means to Evaluate Acceptability of NPP Pittman NRR/DHf5/tQB HICH 12/31/84 Personnel Qualifications Program Hf01.1.4 Review and Evaluate Industry Programs Pittman NRR/DHf5/tQB HIGH 12/31/84 Hf01.2.0 Training - - -
Hf01.2.1 Evaluate Industry Training Pittman NRR/DHF5/LQB HIGH 12/31/84 HF 01. 2. 2 Evaluate INPO Accreditation Program Pittman NRR/DNF5/LQB HIGH 12/31/84 HF01.2.3 Revise Standard Review Plan Section 13.2.3 Pittman N<R/DHf5/tQB HIGH 12/31/84 HF01.3.0 Licensinq Examination - - -
HF01.3.1 Develop Job Knowledge Catalogue Pittman NRR/DHF5/0LB HIGH 12/31/84 HF01.3.2 Develop Licensing Examinations Handbook Pittman NRR/DHF5/0LB HIGH 12/31/84 Hf01.3.3 Develop Criteria for NPP Simulators Pittman NRR/DHF5/DLB HIGH 12/31/84 Hf01.3.4 Iraining Requirements Package (Revise 10 CFR 55 and Pittman NRR/DHF5/0LB HIGH 12/31/84 RCs 1.149 and 1.8)
HF01. 3. 5 Develop Computerized Exam system Pittman NRR/DHF 5/0tB HIGH 12/31/84 Nf01.4.0 Procedures - - -
$ HF 01. 4.1 HF01.4.2 Inspection Module for Upgrading Procedures E0P Effectiveness Evaluation Pittman Pittman NRR/DHF5/PSRB HIGH 12/31/84 NRR/DHF5/PSRB HIGH 12/31/84 Hf01.4.3 Criteria for Safety-Related Operator Actions Pittman NRR/DHF5/PSRB HIGH 12/31/84 Hf01.4.4 Guidelines for Upgrading Other Procedures Pittman NRR/DHf5/PSRB HIGH 12/31/84 Hf 01. 4. 5 Applications of Artificial Intelligence Pittman NRR/DHF 5/PSRB HIGH 12/31/84 HF01.5.0 Man-Machine Interface (MMI) - - -
iiT01.5.1 Local Control Stations Pittman NRR/[WF5/HFEB HIGH 12/31/84 Nf01.5.2 Annunciators Pittman NRR/DHF5/HFEB HIGH 12/31/84 HF01.5.3 Evaluate Operational Aid Systems Pittman NRR/DHF5/HFEB HIGH 12/31/84 HF01.5.4 Computers and Computer Displays Pittman NRR/DHF5/HFEB HIGH 12/31/84 HF01.6.0 Management and Organization - - -
HF01.6.1 Development of Regulatory Position on Management Pittman NRR/DHF5/LQB HIGH 12/31/64 and Organization HF01.6.2 Evaluate Criteria for SALP Reviews Pittman NRR/DHF5/LQB HIGH 12/31/84 HF 01. 6. 3 Revise Standard Review Plan Section 13.1 Pittman NRR/DHF5/LQB HIGH 12/31/84 y HF02 MAINTENANCE AND SURVElltANCE PROGRAM PLAN (MSPP) m -
=
rn 8D o Phase I <
[ Hf02.1.1 Hf02.I.2 Survey Current Maintenance Practices Pittman NRR/DHf5/PSRB HIGH 06/30/85 7*-
s.o Maintenance Perf ormance Indicators Pittman NRR/DHF5/PSRB HIGH 06/30/85 d HF OL ! . 3 HF02.1.4 Monitor Industry Activities Participate in Standards Groups Pittman NRR/DHF5/PSRB HIGH 06/30/85 $
Pittman NRR/DHIS/PSRB HIGH 06/30/85 HF02.1.5 Maintenance and Surveillance Program Integration Pittman NRR/DHf5/PSRB HIGH 06/30/85
- O m
]
O TABLE II (Continued)
N 00 Action lead Lead Office / Safety latest G Plan Item / SPEB Division / Priority latest issuance MPA a Issue No. litle Engineer Branch Ranking Revision Date No.
1, HF02.I.6 Analysis of Japanese /U.S. NPP Maintenance Programs Pittman NRR/DHF5/P5RB HIGH 06/30/85 HF 02. I . / Maintenance Personnel Qualifications Pitt. .i NRR/DHF5/PSRB HIGH 06/30/85 HF02.1.8 Human Factors In In-~5ervice Inspections Pittman NRR/DNF5/PSRB HIGH 06/30/85 HF02.1.9 Human Error in Events Involving Wrong Unit Wrong Pittman NRR/DHf5/PSRB HIGH 06/30/85 Train
.i Phase !!
HF 02. l l . 0 *hase II Tasks To Be Determined Af ter Resolution of Pittman NRR/DHf5/P5RB HIGH 06/30/85 Phase I l
i 1
! W O
1 1
2 C
23 2 rT1 m O <a.
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4 W
1 O N
i @
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I
! TABLE III
SUMMARY
OF THE PRIORITIZATION OF ALL TMI ACTION PLAN ITEMS, i TASK ACTION PLAN IIEMS, NtW GENERIC ISSUES AND HUMAN FACTORS ISSUES tegend NOTES: 1 - Possible Resolution Identified for Evaluation j 2 - Resolution Available j 3 - Resolution Resulted in either the Establishment of New Requirements or No New Requirements j ui 4 - Issues to be Prioritized in the future H $ - Issue that is not a Generic Safety Issue but should be Assigned Resources for Completion
. HIGH - High Safety Priority j MEDIUM - Medium Safety Priority
- LOW - Low Safety Priority
, DROP - Issue Dropped as a Generic Issue 4
USI - Unresolved Safety Issue j I - IMI Action Plan Item with Implementation of Resolution Mandated by NUREG-0737 C
- D l r71 m i in a b 8 <,.
- O W
I W (A,,
l W O
- s i ui i
l O
i cn j
g TABLE Ill (Continued) o N
C' COVERED IN OTHER N0if RESOLVED STAGES NOTE NOTE NOTE NOTt AC110N ITEM / ISSUE GROUP a i ISSUES 1 2 3 U51 HIGH MEDIUM LOW DROP 4 5 TOTAL 1
- 1. THI A" TION PL AN ITEMS (352)
! G: ;afety (i) Generic Safety 94 55 1 1 107 0 8 4 12 7 2 -
291 i (b) Non-Safety Licensing -
0 0 4 51 - - - -
0 0 6 61 (i)
- 2. TASK ACTION PL AN ITEMS (142)
(a) Safety 0 1 14 12 - - - - - -
2/
j (i) USI l (ii) Generic Safety -
18 0 1 28 -- 4 4 3 9 7 74 l 0 0 - - -
0 0 7 i (iii) Regulatory impact -
0 5 1 1 (b) Non-Safety U1 N (i) Licensing -
0 0 0 1 7 0 11 19 l
(ii) Environmental -
1 0 0 6 - - - -
6 0 2 15 I. 3. NEW GENERIC ISSUES (181)
'l
) (a) Safety (i) Generic Safety -
36 4 3 17 0 12 7 5 18 65 -
167 (ii) Regulatory impact -
0 0 0 1 1 0 0 8 10 (b) Non-Safety (i) Licensing -
0 0 0 0 - - - -
0 0 4 4
- 4. HUMAN F AC10R5 ISSUE S (34) z (a) Sare_ty E (i) Generic Safety -
0 0 0 0 0 34 0 0 0 0 -
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Ss 00 cm TABLE IV LI5flNG OF AE00 REPORIS AND RELATED GENERIC ISSUf 5 This listing shows all AEOD reports that have been addressed either as completely new safety issues or as part of new or existing safety issues. It should be noted that, in some cases, more than one AE00 report has been generated on a single topic. However, all AE00 reports related to the identified safety issues are listed alphanumerically including those that have been superseded by other AEOD reports. The following is a description of the types of AE00 reports:
C - Reactor Case Study E - Reactor Engineering Evaluation 5 - Special Study Report I - Technical Review Report
- AE00 Related Related Report Safety ALOD No. AE00 Report Title Issue No. Report On DJ C001 Report on the Browns Ferry 3 Partial failure 41 -
to scram Event on June 28, 1980 C003 Report on Lass of Offsite Power Event at 47 -
Arkansas Nuclear One, Units 1 and 2 C004- AE00 Actions Concerning the Crystal River 3 33 E122 Loss of Non-Nuclear Instrumentation and Integrated Control System Power on February 26, 1980 C005 AE00 Observations and Recommendations Concerning 37, 42 -
the Problem of Steam Generator Overfill and Combined Primary and Secondary Side Blnwdown C101 Report on the Saint Lucie 1 Natural Circulation 31 -
Cooldown on June 11, 1980 C102 H. B. Robinson Reactor Coolant System Leak on 34 -
January 29, 1981 C103 AE00 Safety Concerns Associated with Pipe Breaks 40 -
_g C104 Millstone Unit 2 Loss of 125 V DC Bus Event on 46 -
c: January 2, 1981 gj C105 Report on the Calvert Cliffs Unit i Loss of 36 -
c) Service Water on May 20, 1980 5f ch C201 Safety Concern Associated with Reactor Vessel 50, 101 -
ya Level Instrumentation in Boiling Water Reactors D' OJ -h b) C)
- 3 L@
o cn g I AtsL E IV (Continued) o h
CD AE00 Report Related safety Related ALOD No AE00 Report Title Issue No. Report C202 Report on Service Water System Flow Blockages by 32 E016 Bivalve Mollusks at Arkansas Nuclear One and Brunswick C203 Survey of Valve Operator-Related Events 54 E305 Occurring During 1978, 1979, and 1980 C204 San Onofre Unit 1 Loss of Salt Water Cooling 44 -
Event of March 10. 1980 C205 Abnormal Transient Operating Guidelines (AT0G) 56 -
as Applied to the April 1981 Overfill Event at Arkansas Nuclear One, Unit 1 C301 Failures of Class 1E Safetv-Related Switchgear 55 -
Circuit Breakers to Close on Demand C401 Low Temperature Overpressure Events at Turkey 94 E426 Point Unit 4 C403 Edwin 1. Hatch Unit No. 2 Plant Systems Interaction 85 E322 Event on August 25, 1982 C404 Steam Binding of Auxiliary f eedwater Pumps 93 E325 C501 Safety implications Associated With In-Plant 106 -
Pressurized Gas Storage and Distribution Systems un in Nuclear Power Plants A C503 Decay Heat Removal Problems at U.S. Pressurized 99 -
Water Reactors E002 BWR Jet Pump Integrity ,
12 -
E005 Operational Restrictions for Class IE 120 VAC 48 -
Vital Instrument Buses E007 Potential for Unacceptable Interaction Between 39 -
the Control Rod Drive System and Non-Essential Control Air System at the Browns Ferry Plant E010 Tie Breaker Between Redundant Class IE Buses - 49 -
Point Beach Nuclear Plant, Units 1 and ?
E0ll Concerns Relating to the Integrity of a Polymer 38 -
Coating for Surfaces Inside Containment E016 Flow Blockage in Essential Equipment at ANO 32 C202 Caused by Corbicula sp. (Asiatic Clams)
E101 Degradation of Internal Appurtenances.in LWR Piping 35 -
Ell? Inoperability of Instrumentation Due to Extreme 45 E226
- Cold Weather E E122 AE00 Concern Regarding Ir: advertent Opening of 33 C004 ,
M Atmospheric Dump Valves on B&W Plants During to o loss of ICS/NNI Power 1 m
$ E123 Common Cause failure Potential at Rancho Seco - 43 -
~^
o Desiccant Contamination of Air Lines W E204 Effects of Fire Protection System Actuation on 57 -
S g
Saf ety-Related E quipment w
& O O
_ _ _ _ _ _ _ _ _ . _ _ _ . _ .. - . - - . _ . . ~ _ _ . . _ . . _ . _ . - - _ . . ,._._.__m .m. _-, _ _. . . _ _ _ -_ - - . . ___.m _. .
\ e i
I o i
m i .N g lABLE IV (Continued).
t o l M m
A[OD ' Related Related Report Safety AE00 No. AE00 Report Title issue No. Report i
E209 Generator Rotor Retaining Ring as a Potential 30 -
Missile (Incident at Barseback 1 on 4/13/19) j' E215 Engineering Evaluation of the Salt Service Water 52 . -
System Flow Blockage at the Pilgrim Nuclear.
Power Station by Blue Mussels j E226 Inoperability of Instrumentation Due to Extreme i
45 Ell 2 Cold Weather l
E304 Investigation of Backflow Protection in Common 77 - -
Equipment and floor Drain' Systems to Prevent j Flooding of Vital Equipment in Safety-Related 4
Compartments .,
i E305 Inoperable Motor-Operated Valve Assemblies Due 54 C203
! to Premature Degradation of Motors and/or Improper.
! Limit Switch / Torque Switch Adjustment
- E322 Damage to Vacuum Breaker valves as a Result of Relief 85 C403 i Valve Lifting
- E325
' Vapor Binding of Auxiliary Feedwater Pumps a' 93 C404 Robinson 2 l
E414 Stuck Open Isolation Check Valve on the Residual ~105 -
m Heat Removal
i Leading to Primary to Secondary Side Leakage E426 Single Failure Vulnerability of Power Operated 94 C401 Relief Valve (PORV) Actuation Circuitry for Low
- Temperature Overpressure Protection (LTOP)
- S401 Human Error in Events Involving Wrong Unit or 102 -
Wrong Train 1302 Postulated Loss of Auxiliary Feedwater System 4 68 -
j Resulting from a Turbine Driven Auxiliary j
Feedwater Pump Steam Supply Line Rupture T305 Flow Blockage in Essential Raw Cooling Water
- 51 -
System Due to Asiatic Clam Instrusion at Sequoyah 1
) T420 Failure of an Isolation Valve of the Reactor Core 81 -
) Isolation Cooling System to Open Against Operating
- Reactor Pressure 2
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Rsvision 1
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TASK III.D.3: WORKER RADIATION PROTECTION IMPROVEMENT The objective of this task is to improve nuclear power plant worker radiation protection to allow workers to take effective action to control the course and consequences of an accident, as well as to keep exposures as low as reasonably achievable (ALARA) during normal operation and accidents, by improving radiation protection plans, health physics, inplant radiation monitoring, control room habitability, and radiation worker exposure data base.
ITEM III.D.3.1: RADIATION PROTECTION PLANS DESCRIPTION Historical Background The purpose of this TMI Action Plan 48 item is to improve nuclear power plant worker radiation protection programs by better defining the criteria and respon-sibility for such programs. Detailed appraisals of health physics programs at all operating nuclear power plants were performed in 1980 and 1981. These appraisals, summarized in NUREG-0855,204 indicated that certain generic deficien-cies existed at many plants due in part to lack of specific performance criteria p
g and/or assigned responsibility for programs. The establishment of a radiation protection plan as a guiding document for implementing procedures has been b proposed as a method for formalizing commitment to specific performance criteria contained in Regulatory Guides and SRP Section 12.11 Proposed guidance and acceptance criteria for radiation protection plans have been published in draft form as NUREG-0761.20s A proposed amendment to 10 CFR 50 has been drafted.20s Safety Significance The development of radiation protection plans has no impact on public safety.
Instead, the safety significance lies in the reduction of occupational exposure.
Possible Solutions As currently envisioned, radiation protection plans would tie together specific implementing procedures, many of which currently exist at licensed plants.
Additional procedures may be required at many plants to fully implement the
~
plan; however, extensive revision of procedures should not generally be required.
Administrative and technical manpower would be required to develop the plan, revise and write procedures as necessary, and some additional equipment (such as additional survey equipment) may be required. Installation of such equip-ment should not require any significant work in radiation areas. The benefit of radiation protection plans would be in two primary areas: (1) reduction of individual and collective dose due to improved planning and controls for work in radiation areas, and (2) improved confidence in results of radiation pro-tection programs.
o 06/30/86 1.III.D.3-1 NUREG-0933
Revision 1 PRIORITY DETERMINATION The assessment of this issue and its resolution was first performed 64 by con-sensus opinions of four PNL health physicists who were extensively involved in the Health Physics Appraisal Program. These personnel included expertise from both industry and regulatory sides of the issue. Estimates of routine cost and probable man-rem reductions were discussed and agreed upon. For core-melt accident recovery and refurbishing, the panel assumed man-rem savings comparable on a percentage basis to those for routine operations. The cost impact of these man-rem savings was then estimated by a PNL expert involved in estimating accident recovery costs.
Frequency / Consequence Estimate m
There are three terms in the estimation of occupational dose change due to this safety issue. .These are the change due to accidents, the change due to issue resolution implementation, and the change due to resolution operation.
The estimated change due to accidents (the first term) is the change in the product of accident frequency and occupational dose associated with the recovery from an accident. As previously stated, no change in accident frequency is expected to occur due to this issue. However, a small change in occupational accident recovery dose is expected. Radiation protection plans are primarily oriented toward routine plant operation. In the event of a major core-melt accident, specialized procedures would have to be developed. Having the upgraded radiation protection plan for normal operation in place, however, is expected to result in improved specialized procedures if required. The resulting reduc-tion in occupational dose for plant recovery is estimated to be slightly less than 5%. Using the estimates of total occupational dose resulting from recovery from an accident, as listed in Appendix D of NUREG/CR-2800,64 this works out to 3.3 x 10 2 man-rem /RY for BWRs and 7.4 x 10 2 man-rem /RY for PWRs.
The implementation of radiation protection plans (the second term) would be an administrative effort. Therefore, there is zero exposure associated with implementation.
The establishment of radiation protection plans is estimated to result in a reduction of occupational risk during operation (the third term). This reduc-tion would be due to improved controls on personnel dose and an improved ALARA Program. PNL's experts estimated the occupational dose reduction to be on the order of 5%.64 However, the Occupational Radiation Protection Branch ~(0RPBR) of RES has been investigating the costs cnd benefits associated with radiation protection plans. Based on a comparison of plants with and without major radiation protection plans, it was estimated that occupational doses could be reduced by at least 10%. Savings of 25% or more appear achievable.207 The 1980 average occupational dose was about 800 man-rem. Therefore, we will assume that radiation protection plans could avert 200 man-rem /RY.
Cost Estimate PNL estimated that 35 man-weeks at a cost of $35,000 and equipment worth
$50,000 would be required per plant to implement the radiation protection plans.64 In order to operate under the new radiation protection plans, it was 06/30/86 1.III.b.3-2 NUREG-0933
Revision 1 O
/ felt that most plants would have to add personnel. It was estimated that one pro-fes'sional and one technical staff member would be needed. At 52 weeks per year, this gives an additional 104 man-weeks per year for each plant, or'$104,000 plant cost per year.
However, ORPBR has noted that the licensees' cost will vary widely depending on the adequacy of the present program.20s In addition, since radiation protec-tion plans have the effect of reducing the time workers are exposed, indiv'idual tasks are often speeded up. Some licensees have found that the savings resulting from reduced downtime have compensated for the cost of the program.
Currently, there are 43 operating PWRs with a cumulative experience of 350 RY and 27 BWRs with a cumulative experience of 260 RY. If we add to these the 36 PWRs and 21 BWRs under construction and assume a plant lifetime of 30 years, there are~3,200 RY remaining: 1,180 RY for BWRs and 2,020 RY for PWRs.
ORPBR has estimated that 5 NRC staff years will be required.20s Thus, NRC costs are estimated to be $500,000.
The total cost associated with the solution to this issue is $340.5M.
Value/ Impact Assessment The total risk reduction associated with this issue is 6.4 x 105 man-rem.
Therefore, the value/ impact score is given by: ,
[ 3 ,6.4 x 105
$340.5M man-rem
= 1,880 man-rem /$M Uncertainties The dominant parameters in the evaluation of this issue are the percent saving in occupational dose during normal operation, which is unlikely to be incorrect by more than a factor of ten, and the cost to the licensee, which is expected to be within a factor of 5. This implies a range in 5 from 100 to 30,000 man-rem /$M and a range in total man-rem saved of 6 x 104 to 6 x 106
-CONCLUSION Based on the value/ impact score and potential reduction in occupational dose, this issue was give a high priority ranking. In resolving this issue, the staff agreed to support alternative regulatory concepts which recognize the contribu-tions of industry self policing programs to the extent that such programs are ef fective and consistent with NRC regulatory responsibilities. As a result, the staff entered into a " Coordination Plan for Radiological Protection Activ-ities" with INP0 under a " Memorandum of Agreement Between INPO and the USNRC."
Under this agreement, over the two year period outlined in the Coordination Plan, NRR staff developed a method for evaluating industry performance in radia-tion protection programs incorporating ALARA concepts at power reactors and p observed the INP0 evaluation and assistance process at a number of operating facilities.
06/30/86 1.III.D.3-3 NUREG-0933
R: vision 1 The staff performed analyses of a number of radiological data trends as part of the effort to determine if-the power reactor industry has achieved successful ALARA-integrated radiation protection programs. An analysis of these trends and portions of the supporting data bases were documented in the report, " Summary Analysis of Selected Radiological Trends at Power Reactors."912 Following the staff's compilation of data and evaluation of a number of trends in radiological protection at power reactors, the staff concluded that most power reactor radiation protection programs are adequately incorporating ALARA concepts and can satisfactorily perform at a level which meets the objectives of Item III.D.3.1 Thus, this issue was RESOLVED and no new requirements were established.913 ITEM III.D.3.2: HEALTH PHYSICS IMPROVEMENTS The four parts of this item have been combined and evaluated together.
DESCRIPTION Historical Background In this TMI Action Plan 48 item, four specific items were identified for resolu-tion: (1) Requirement for Use of Certified Personnel Dosimeter Processors; (2) Audible Alarm Dosimeter Regulatory Guide; (3) Develop Standard Performance Criteria for Radiation Survey and Monitoring Instruments; and (4) Develop Air Purifying Respirator Radioiodine Cartridge Testing and Certification Criteria.
Item (2) will not be considered further since a regulatory guide has been issued in final form prior to this evaluation. Thus, Item (2) is considered reso'lved.
Safety Significance (1) Requirement for Use of Certified Personnel Dosimetry Processors The proposed resolution would amend 10 CFR 20 to require that only nationally certified dosimetry-processors be used by NRC licensees for personnel radi-ation dosimetry. Processors would be required to meet ANSI N13.11 (or its replacement standard) criteria for testing. Certification of processors would be performed by the National Voluntary Laboratory Accreditation Program (NVLAP) administered under the auspices of the U.S. Department of Commerce (DOC).
This issue does not specifically address core-melt accidents nor the public risk, occupational dose, or accident avoidance costs associated with such accidents. It is related to the worker's right to accurate measurements of occupational dose. The proposed resolution would require accurate and precise determinations of individual worker doses using dosimeters, readout systems, and processing procedures certified to be capable of meeting minimum criteria defined in a national standard. The administrative and regulatory limits for occupational dose would be unaffected by this work.
A draft ANSI standard (ANSI N13.11) for dosimeter testing was issued for trial use in 1978. This standard has undergone substantial testing and NUREG-0933 06/30/86 1.III.D.3-4 l
Revision 1 t
O remains only to be finalized for issuance as a new ANSI standard. Once U/ issued, it will form the basis for amending 10 CFR 20. Testing and certifica-tion of dosimeter processors for criteria contained in this standard will be performed by NVLAP under D0C.
(2) This item has been resolved as discussed before.
2 (3) Develop Standard Performance Criteria for Radiation Survey and Monitoring Instruments Testing of radiation survey and monitoring instruments will provide a high degree of quality assurance that instruments are capable of performing intended functions under specified conditions. This will allow consistent utilization of workers without impacting current individual or collective occupational dose. A draft' standard for health physics instrumentation testing (ANSI N42.17-D2) has been developed.
This standard will undergo a. field trial period, using off-the-shelf instruments,-to determine its adequacy. This trial period is presently l estimated to continue through FY-1984 and is jointly funded by NRC and the Department of Energy (DOE) at $400,000 each. Following the trial period, a final standard will be adopted by NRC and only those instruments meeting this standard would be ac'eptable c for use in NRC licensed facilities.
At this time, a plan for implementing the testing program has not beerf developed. It is anticipated, however, that independent testing laboratories f,s 'would, for a fee, test instruments submitted by vendors or reactor licensees.
The testing laboratories would be certified by NVLAP under D0C. Costs i'
associated with NVLAP certification and instrument testing. fees would be '
4 .
passed on to industry in the form of higher instrument prices.
(4) Develop Air Purifying Respirator Radiciodine Cartridge Testing and Certification Criteria -
Air purifying respirators are not currently acceptable for radioiodine pro-2 tection due to the lack of accepted test procedures for certifying cartridge filtering efficiency. The result is that bulky self-contained breathing apparatus (SCBA) must be worn by workers in radiciodine environments. Such environments are expected during and after core-melt accidents. The results of wearing SCBA is to substantially reduce worker efficiency due to physical stress and the relatively short working time limited by air tank capacity.
Use of air purifying respirators would reduce worker stress and improve worker efficiency.
It is expected that operator dose would be unaffected by the availability of respirators. Immediately after an accident, SCBA would still be used 4
due to immediate hazards. During long-term recovery activities respirators 3 could be used. However, reduced external dose due to efficient use of time in radiation zones is expected to be offset by the reduced effectiveness of the respirators, compared to SCBA, in avoiding internal exposures. Criteria i
and test procedures for radioiodine cartridges have been under development by LASL using NRC funds. The technology has been developed and is in the O process of being transferred to NIOSH. When transfer is complete, it is anticipated that NIOSH will amend 30 CFR 11 to incorporate the testing 06/30/86 1.III.D.3-5 NUREG-0933
Revision 1 methods and criteria into respirator test and certification schedules.
Respirator and cartridge manufacturers would submit products for certi-fication testing and periodic quality control checks would be performed.
Following establishment of certification programs, NRC evaluation is antic-ipated regarding the need to specify the quantity and types of respirators necessary for normal and emergency use at a typical power reactor.
This issue will have no impact on public risk associated with core-melt accidents. The occupational dose impact is also considered to be zero, the benefit to workers being reduced stress, improved comfort and, conse-quently, better worker performance.
CONCLUSION The issues and their proposed resolutions do not impact public risk nor are they expected to increase or decrease occupational dose. They relate to tne rights of workers to be assured of adequate radiation protection and would reduce stress during the performance of work in radiation zones. Therefore, this item is considered to be a Licensing Issue.
ITEM III.D.3.2(1): AMEND 10 CFR 20 This Licensing Issue was evaluated in Item III.D.3.2 above and its solution is available.
ITEM III.D.3.2(2): ISSUE A REGULATORY GUIDE O
This Licensing Issue was evaluated in Item III.D.3.2 above and was determined to be resolved.
ITEM III.D.3.2(3): DEVELOP STANDARD PERFORMANCE CRITERIA This Licensing Issue was evaluated in Item III.D.3.2 above and its solution is available.
ITEM III.D.3.2(4): DEVELOP METHOD FOR'~TC571NG AND CERTIFYING AIR-PURIFYING RESPIRATORS This Licensing Issue was evaluated in Item III.0.3.2 above and its solution is available.
ITEM III.D.3.5: RADIATION WORKER EXPOSURE The three parts of this item have been combined and evaluated together.
DESCRIPTION This TMI Action Plan 48 item called for the NRC to continue its efforts to improve and expand the data base on industry employees in order to facilitate possible 06/30/86 1.III.D.3-6 NUREG-0933
A Revision 1 O
V future epidemiological studies on worker health.
are as follows:
The three parts of this item (1) " Improve and expand the data base on industry employees." This item is considered important in improving a data base used by the NRC in
! judging the adequacy of its radiation protection standards. Meetings have been held with DOE, ORM, NCI, AIF, and officials of Canadian and British national dose registries and health statistics organizations to discuss issues related to this item. Although these meetings have resolved certain generic issues, this item is a long-term goal requir-ing on going cooperation between nuclear regulators, industries, and workers.409 (2) " Investigate non-legislative means of obtaining employee health data."
- This item was completed in September 1982 following discussions, about worker health data with DOE, AIF, EPRI, and officials of British and Canadian national dose registries and health statistics organizations.409 (3) " Include as part of the overall rewrite of 10 CFR Part 20 considera-tion of a requirement for licensees to collect worker medical data."
This item was completed in February 1981 following a decision by the Part 20 task force not to require the collection of worker medical data.409 The value of this item does not lie in the reduction of public or occupational risk. Instead, it will provide data on which future regulatory decisions will be based. Therefore, this item is not directly related to public safety and
( is considered a licensing issue.
CONCLUSION This item is a Licensing Issue.
ITEM III.D.3.5(1): DEVELOP FORMAT FOR DATA TO BE COLLECTED BY UTILITIES REGARDING TOTAL RADIATION EXPOSURE TO WORKERS This item was evaluated in Item III.D.3.5 above and was determined to be a Licensing Issue.
ITEM III.D.3.5(2): INVESTIGATIVE METHODS OF OBTAINING EMPLOYEE HEALTH DATA BY NONLEGISLATIVE MEANS This Licensing Issue was evaluated in Item III.D.3.5 above and was determined to be resolved.
ITEM III.D.3.5(3): REVISE 10 CFR 20 This Licensing Issue was evaluated in Item III.D.3.5 above and was determined to be resolved.
06/30/86 1.III.D.3-7 NUREG-0933
Revision 1 REFERENCES
- 11. NUREG-0800, " Standard Review Plan," U.S. Nuclear Regulatory Commission.
- 48. NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident,"
U.S. Nuclear Regulatory Commission, May 1980.
202. Memorandum for G. Cunningham, et al., from K. Goller, " Proposed Amendment to Part 50 on Radiation Programs, Including ALARA," September 10, 1982.
204. NUREG-0855, " Health Physics Appraisal Program," U.S. Nuclear Regulatory Commission, March 1982.
205. NUREG-0761 " Radiation Protection Plans for Nuclear Power Reactor Licensees," U.S. Nuclear Regulatory Commission, March 1981.
206. Memorandum for L. Rubinstein from M. Ernst, " Proposed Position Regarding Containment Purge / Vent Systems," April 17, 1981.
409. Memorandum for W. Minners from W. Mills, "Prioritization of Generic Issue III.D.3.5, Radiation Worker Data Base," February 22, 1983.
912. Memorandum to T. Murley, et al. , from H. Denton, " Evaluation of Industry Success in Achieving ALARA-Integrated Radiation Protection Plans - Data Trend Assessments," May 19, 1986.
913. Memorandum for V. Stello from H. Denton, " Resolution of Generic Issue III.D.3.1, ' Radiation Protection Plans,'" May 19, 1986.
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U ITEM C-4: STATISTICAL METHODS FOR ECCS ANALYSIS DESCRIPTION Historical Background As identified in NUREG-0471,3 Appendix K of 10 CFR 50 specifies the requirements for LWR ECCS analysis. These requirements call for specific conservatisms to
'be applied to certain models and correlations used in the analysis to account for data uncertainties at the time Appendix K was written. This issue involved the staff development of.a statistical assessment of the certainty level of the 2200 F peak cladding temperature limit.
Safety Significance The purpose of this issue was to aid the staff in the review of changes to vendor ECCS models and in the performance of staff audit calculations of ECCS performance. Therefore, no significant change in.public risk is attributed to this issue.
CONCLUSION O In accordance with a DSI memorandum,907 the staff informed the Commission in
() SECY-83-472908 of their intent to allow voluntary use of statistical uncertainty analysis by the industry to justify relaxation of all but the required conser-vatisms contained in current ECCS evaluation models. In addition, the staff is currently preparing an ECCS rule change that will allow the use of a best esti-mate model plus a> statistical uncertainty to determine the peak cladding tem-perature. Until the staff revisions to the Appendix K rule change are imple-mented, the staff proposed to accept the revised ECCS analysis methods for demonstrating conformance to~the current Appendix K requirements. Thus, this item was determined to be a resolved Regulatory Impact issue.
REFERENCES
- 3. NUREG-0471, " Generic Task Problem Description (Categories B, C, and D),"
U.S. Nuclear Regulatory Commission, June 1978.
907. Memorandum for W. Minners from 8. Sheron, " Generic Issues C-4, C-5, C-6,"
May 29, 1985.
908. SECY-83-472, " Emergency Core Cooling System Analysis Methods," November 17, 1983.
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ITEM C-5: DECAY HEAT UPDATE DESCRIPTION Historical Background l
As identified in NUREG-04713, this issue involved following the work of research groups in determining best estimate decay ~ heat data and associated uncertainties for use in LOCA calculations.
Safety Significance l No significant change in public risk is attributed to this issue.
CONCLUSION In accordance with a OSI memorandum,007 the staff has determined that the 1979 ANSI /ANS Standard 5.1920 is technically acceptable and has allowed the use of i
~this data to justify relaxation of non-required conservatisms in current ECCS '
evaluation models.808 In addition, the proposed ECCS rule change being developed by the staff will allow the use of this new data. Thus, this issue is a resolved Regulatory Impact issue.
' b REFERENCES
- 3. NUREG-0471, " Generic Task Problem Descriptions (Categories B, C and D),"
U.S. Nuclear Regulatory Commission, June 1978.
907. Memorandum for W. Minners from B. Sheron, " Generic Issues C-4, C-5, C-6,"
May 29, 1985.
908. SECY-83-472, " Emergency Core Cooling System Analysis Methods,"
November 17, 1983.
920. ANSI /ANS 5.1, " Decay Heat Power in Light Water Reactors," American National Standards Institute, 1979.
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! ITEM C-6: LOCA HEAT SOURCES DESCRIPTION Historical Background As identified in NUREG-0471,3 this issue involved staff-evaluations of vendors' data and approaches for determining LOCA~ heat sources and developing staff positions as needed. The contributors to LOCA heat scurces, along with their associated uncertainties and the manner in which they are combined, have an impact on LOCA calculations.
The staff evaluated the combined effect of power density, decay heat, stored energy, fission power decay, and their associated uncertainties with regard to calcuations of LOCA heat sources.
4 Safety Significance No significant change in public risk is attributed to this issue.
CONCLUSION In accordance with a OSI memorandum,807 the staff informed the Commission 90s
-of their intent to allow the statistical combination of heat sources to justify the relaxation of non-required conservatisms in ECCS evaluation models. Also, the proposed ECCS rule change being developed by the staff would allow the statistical combination cf LOCA heat sources. Thus, this issue is a resolved Regulatory Impact issue.
REFERENCES
- 3. NUREG-0471, " Generic Task Problem Descriptions (Categories B, C, and D),"
U.S. Nuclear Regulatory Commission, June 1978.
907. Memorandum for W. Minners from B. Sheron, " Generic Issues C-4, C-5, C-6,"
i May 29, 1985.
908. SECY-83-472, " Emergency Core Cooling System Analysis Methods,"
November 17, 1983.
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'wJ ISSUE 3: SET POINT DRIFT IN INSTRUMENTATION DESCRIPTION Historical Background This issue is identified in Appendix D of NUREG-05724 and is one of the key observations made after the ACRS requested its members and consultants to make comprehensive reviews of all LERs issued during the years 1976, 1977, and 1978.
Data collected over this 3 year period showed that 10% of all LERs were related to drift in the set points of instrumentation beyond Technical Specification (TS) limits. This amounted to an average of 258 LERs for each of the 3 years considered.
Safety Significance An unplanned change in the set point of an instrument (set point drift) will alter the actual value of the measured parameter at which a particular action is to occur. Drift in the set point of an instrument could result in delay in the initiation of a safety function.
Possible Solutions The solutions proposed in NUREG-05724 were as follows: (1) where set point
\ drift is due to component failures, make the necessary repair, recalibrate, and restore the instruments to service, and (2) increase the margin between the selected set points and TS limits in order to accommodate inherent instrument inaccuracies that result in' set point drift.
CONCLUSION Because of the qualitative judgment that the safety significance is very small and, because of the difficulty in trying to quantify the risk importance of this issue due to the fact that the drift problem would permeate all of the accident analyses, no attempt was made at quantification.
While set point drif t might result in some small degradation of safety function,
'it would not prevent the initiation of a safety function or result in a major degradation because of the following reasons: (1) safety instruments are redun-dant-and diverse, (2) the measured variables change rapidly, and (3) drift is periodically detected and corrected. Since all safety channels are redundant, the instruments in both channels would have to drift at the same time in order for the delay to occur. But the redundant instruments are identical, sense the same variable, and operate under the same conditions and, therefore, the amount of drift would not be independent. The number of LERs reporting simultaneous drift in redundant channels was not given in NUREG-0572.4 Most safety functions are initiated by diverse signals emitted from different instruments. Drift in any one of these instruments would be expected to be
("j independent of drift in the remaining instruments. Therefore, instrument drift would be expected in many (but not all) cases to result in some delay in the initiation of a safety function.
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Revision 1 Most of the process variables used to initiate safety systems have high rates of change during accidents that would result in significant consequences.
Therefore, the time delay in reaching an instrument trip point is insensitive to set point.
Drift is detected and corrected during regular surveillance of safety instru-mentation. Thus, the degree of drift is limited. Repair and recalibration of instruments that have drifted beyond TS limits are actions that are required of licensees in order to keep their plants in operation. In fact, the strict NRC reporting requirements for these events produced the large volume of LERs that was the basis for consideration of instrumentation set point drift as a generic safety issue. To some degree, this reporting requirement is an incentive to licensees to reduce or allow for instrument set point drift.
In order to address the establishment and maintenance of set points for indi-vidual safety-related instrument channels, the Instrument Society of America (ISA) issued Draf t F to ISA 567.04" on May 22,1979. This standard was adopted by the NRC and was incorporated in the Proposed Revision 2 to Regula-tory Guide 1.10550 which was issued for public comment in December 1981. In February 1986, the staff published Revision 2 to Regulatory Guide 1.105 endors-ing the guidance of ISA 567.04-1982. No revisions to the SRP11 were required.
Thus, this issue was RESOLVED and no new requirements were established.903 REFERENCES
- 4. NUREG-0572, " Review of Licensee Event Reports (1976-1978)," U.S. Nuclear Regulatory Commission, September 1979.
- 11. NUREG-0800, " Standard Review Plan," U.S. Nuclear Regulatory Commission.
- 49. ISA 567.04 (ANSI N719), Draft F, "Setpoints for Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants," Instrument Society of America, May 22, 1979.
- 50. Draft Regulatory Guide and Value/ Impact Statement, TASK IC 010-5, " Proposed Revision 2 to Regulatory Guide 1.105, Instrument Setpoints," U.S. Nuclear Regulatory Commission, December 1981.
903. Memorandum for T. Speis from H. Denton, " Resolution of Generic Issue 3,
'Setpoint Drift in Instrumentation,'" May 19, 1986.
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V ISSUE 21: VIBRATION QUALIFICATION OF EQUIPMENT l l
DESCRIPTION This issue originated from NUREG-05724 where the ACRS identified vibration-induced equipment failures resulting from various loading conditions. It was also included in the initial list 2s of issues that were identified for prioriti-zation. Since-then, the staff has addressed vibration-induced equipment fail-ures resulting from a number of sources, e.g. , flow-induced vibration (Issue 7),
dynamic water hammer loads (USI A-1), safety relief valve loads (USI A-8),
asymmetric loads (USI A-2), and seismic loads.
The seismic loads and their resulting vibration effects on equipment fragility and operability are included in the following issues: USI A-40, " Seismic Design Criteria - Short-Term Program"; USI A-41, "Long Term Seismic Program"; USI A-46, '
" Seismic Qualification of Equipment in Operating Plants"; B-24, " Seismic Qualifi-cation of Electrical and Mechanical Components"; and B-52, " Fuel Assembly Seis-mic and LOCA Responses." Other issues that address potential vibration-induced problems include C-12, " Primary System Vibration Assessment," and B-73, " Monitor-ing for Excessive Vibration Inside the Reactor Pressure Vessel."
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As evident from above, vibration-related problems can arise from many sources (V and no single, simple, overall solution may be appropriate for the broad spectrum of potential vibration problems. However, since the ACRS identified this issue, the staff has revised the vibration qualification guidance in SRP11 Sections 3.9.2 and 3.10 (see Issue B-24). In addition, the above-identified issues address specific vibration problems that the staff is handling on a case-by-case basis.
CONCLUSION Based on the actions identified above, the original ACRS concerns have been addressed by the SRP revisions and the related issues that are resolved, nearly resolved, or will be resolved on a case-by-case basis. Therefore, we recommend that this issue be DROPPED as a separate issue.
REFERENCES
- 4. NUREG-0572, " Review of Licensee Event Reports (1976-1978)," U.S. Nuclear Regulatory Commission, September 1979.
- 11. NUREG-0800, " Standard Review Plan," U.S. Nuclear Regulatory Commission.
- 23. Memorandum for W. Minners from K. Kniel, " Generic Issues List,"
September 18, 1981.
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ISSUE 36: LOSS OF SERVICE WATER DESCRIPTION Historical Background On June 19, 1981, AE00. issued a preliminary report.,s7 for review and comment on their case study of the incident at the Calvert Cliffs Unit 1 plant in which the plant lost both redundant trains of service water when the system became air bound. On August 5, 1981, NRR provided comments 494 on this report, but indicated that their review was limited because of'the preliminary nature of the AEOD report which did not contain recommendations and conclusions. Subsequently, AE00 issued the final versionssa of this case study on December 17, 1981, in which three generic recommendations were presented in the cover letter and NRR was requested to comment on this version of the AE00 work. NRR provided another reviews 59 of the case study on September 23, 1982, which was more detailed and specifically addressed the AE0D generic recommendations. However, in an AEOD memorandums 88 to NRR on May 2,-1983, additional clarification to the case study was provided. In response to this memorandum, DL requested 481 DSI to review the latest AE0D memorandum. DSI performed this review and provided specific and detailed responsesss2 to all concerns identified by AE0D. Based on the content of the DSI memorandum,ss2 the AE0D concerns were addressed and solutions were b.
identified in a memorandums 83 from NRR to AE0D on September 15, 1983. Also, in an AE00 memorandumss4 to OIE on August 18, 1983, detailed information was pro-
~v ided so that an appropriate Information Notice could be issued by OIE. As a result, IE Information Notice No. 83-77585 was issued on November 14, 1983.
Safety Significance Calvert Cliffs Unit 1 experienced a loss of both redundant trains of service water when the system became air-bound as a result of the failure of a non-safety-related instrument air compressor aftercooler. The significance of this event lies in the fact that it involved two fundamental aspects in the design of safety-related systems: (1) interaction between safety and non-safety-related systems and components, and (2) common cause failure of redundant safety systems.
Possible Solutions A summary of the AEOD recommendations and the NRR responses are as follows:
- I. AE00 Recommendations of Section 8, Part (a) of the AE00 Reportss8
- 1. AE0D Recommendation (1) i It is recommended that butterfly valves SW-4 and SW-5 have valve operators added (pneumatic or electric motor) and that these valves either close automatically, as do the valves on the turbine supply n header, or as a minimum have the capability to be remote manually-operated from the control room.
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R vision 2 NRR Responses 559'583 Inasmuch as the service water system at Calvert Cliffs is a " dual purpose" system, as defined in SRP11 3.6.1, a single failure in one redundant safety-related train is not postulated concurrently with a moderate-energy line break in the other train. Therefore, it is always assumed that the check valve will function as designed to provide the isolation of both trains. Consequently, this configuration is accept-able under the current staff criteria.
However, during the review of operating reactor compliance with the moderate energy pipe break criteria, the staff considered the effect of leak rate from postulated cracks on the system operability and the time required to isolate the break before loss of system safety func-tion. Based on these reviews, the NRR staff believes that sufficient instrumentation (service water head tank level alarms) and time (30 minutes) are available for the Calvert Cliffs operators to locally close the manual isolation valve in such an event. Further, makeup to the heat tank provides additional time for taking this action.
- 2. AE0D Recommendation (2)
It is recommended that the four check valves and the four solenoid operated three-way valves in the instrument air lines that provide control air for the four diesel generator 12 service water supply and return valves be added to the IST program.
NRR Response ss9 A number of plant-specific recommendations were made for the Calvert Cliffs plant. Although NRR agrees that implementation of the AE0D plant-specific recommendations may be beneficial to the operation of the facility, NRR does not believe that ordering such changes would be accompanied by an appropriate increase to the health and safety of the public. Considering that the licensee participated in the peer review of the AE0D case study, they are aware of the AE00 plant-specific recommendations. Therefore, NRR does not believe that further regulatory actions are necessary.
NRR does, however, have the following comment on this plant-specific recommendation. The NRR/0IE working group on instrbment and service air system, which was formed after growing concerns of air system degradation, was considering generic recommendations regarding the isolation and boundary between safety and non-safety-related air systems. By copy of the NRR memorandum,559 the AE0D recommendation to include system boundary valve in the IST program was forwarded to the working group for their information. This working group was dis-banded before they could review this recommendation. Therefore, a memorandums 88 was issued to DL for their review and action on this item.
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(G 3. AE0D Recommendation (3)
If there is a loss of offsite power.and the service water system supplying diesel generator 12 becomes unavailable, the diesel will transfer to an inactive service water loop. Since operator action is necessary to realign the diesel generator such that it energizes the bus which powers the service water pump in the loop to which it was transferred or, alternatively, to start a third service water pump, it is recommended that the human factors of these actions be evaluated against the length of time the diesel can run without service water before it trips.
NRR Responsessa In general, NRR believes that implementation of AE0D's plant-specific recommendations may be beneficial to the operation of the facility.
Therefore, under a separate letter,567 NRR has forwarded a copy of the AE0D final report to the licensee with a statement that NRR considers the recommendations to be valid and that implementation of the AE00 plant-specific recommendations be considered by the licensee.
- 4. AE0D Recommendations (4) and (6)
For operating plants and plants currently in the licensing process that have service water systems that contain both safety and non-safety-related portions, it is recommended that the system isolation (O/
provisions be reviewed to identify any procedural or hardware changes
-necessary to protect the safety-related portion of the service water system from a failure portion of the service water system from a failure in the non-safety-related portion during normal operation and accident conditions.
It is recommended that the guidance in the SRP11 be clarified to.empha-size automatic isolation of the non-safety-related portion of the service water system when it degrades the operability of the safety-related portion of the system.
NRR Responses 559'563 NRR concurs that the guidelines in SRP11 Section 9.2.2 could more clearly stated when automatic isolation of safety and non-safety-related portions of the system is necessary, such as indication of low pressure in the non-safety-related portion as would occur in a failure of the non-seismic Category I piping due to an SSE. The staff will propose revisions to the review procedures (Paragraph III.3.a) of SRPti Section 9.2.2 to more clearly indicate the types of isolation signals required and when isolation is necessary.
- 5. AE0D Recommendation (5) p It is recommended that an IE Circular on common cause failure of service water systems be issued.
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06/30/86 3.36-3 NUREG-0933
Revision 2 NRR Response An IE Information Notice 585 was issued on November 14, 1983.
II. AE0D Recommendations of Section 8, Part (b) of the AE00 Report S58 All of the recommendations (1-3) in this section concern SGTR events. Upon approval by CRGR and the Commission, the action on these recommendations will be carried forward under Steam Generator Staff Actions (Issue 67),
stated in References 559 and 563, as the generic studies being performed by the staff.
III. Additional AEOD Recommendation 580 AEOD Recommendation The accessibility of the steam generator dump valves (ADVs) during accident conditions be reviewed to determine the acceptability of the assumption that the affected steam generator can be isolated in 30 minutes with manual operation of the ADVs.
NRR Response 583 This matter will also be adoressed under I-ssue 67, " Steam Generators Staff Actions" subsequent to approval. However, it is noted that there are cur-rently no plans for the backfitting of RSB BTP 5-1. A cost / benefit analy-sis concerning backfit of RSB BTP 5-1 is now implicitly a part of USI A-45 which will address decay heat removal system improvements including con-sideration of the ADV. The staff requirements for the successful comple-tion of this effort are outlined in a DST memorandum.588 CONCLUSION Based on the contents of References 563 and 564, all but one generic concern and one plant-specific matter raised by the AE0D case study on the Calvert Cliffs loss of service water have been or will be adequately addressed as part of USI A-45 or Issue 67.5.2. The generic concern was resolved with the issuance of SRPti Sections 9.2.1, Rev. 4, and 9.2.2, Rev. 3, in June 1986.901 These revi-sions did not incorporate any new guidelines or requirements.<ao2 The remaining plant-specific matter concerning Calvert Cliffs has been brought to the atten-tien of the DL for appropriate action.588 Thus, this issue has been RESOLVED and no new requirements were established.
REFERENCES
- 11. NUREG-0800, " Standard Review Plan," U.S. Nuclear Regulatory Commission.
494. Memorandum for C. Michelson from H. Denton, "AEOD Preliminary Report on Calvert Clif fs Unit 1 l oss of Service Water," August 5,1981.
557. Memorandum for M. Denton and V. Stello from C. Michelson, "Calvert Cliffs Unit 1 Loss of Service Water," June 19, 1981.
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Rsvision 2 s,) 558. Memorandum for H. Denton and R. DeYoung, Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980," December 17, 1981.
559. Memorandum for C. Michelson from H. Denton, "NRR Comments on AEOD Final Report: Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980,"
September 23, 1982.
560. Memorandum for H. Denton from C. Heltemes, " Response to NRR Comments on AE0D Report, 'Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980,'"
May 2, 1983.
561. Memorandum for W. Houston and L. Rubenstein from F. Miraglia,." Response to NRR Comments on AE0D Report 'Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980'," June 2, 1983.
562. Memorandum for F. Miraglia from W. Houston and L. Rubenstein, " Comments to AE00 Memo dated May 2, 1983, on Calvert Cliffs, Unit 1, Loss of Service Water on May 20, 1980," July 22, 1983.
563. Memorandum for C. Heltemes from H. Denton, " Response to NRR Comments on AE00 Report, Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980,"
September 15, 1983.
564. Memorandum for R. Baer from K. Seyfrit, " Case Study, 'Calvert Cliffs Unit 1 p) Loss of Service Water on May 29, 1980,'" August 18, 1983.
4
(/ 565. IE Information Notice No. 83-77, " Air / Gas Entrainment Events Resulting in System Failures," U.S. Nuclear Regulatory Commission, November 14, 1983.
566. Memorandum for G. Holahan from W. Minners, "Prioritization of Issue 36:
Loss of Service Water at Calvert Cliffs Unit 1," November 10, 1983.
567. Letter to A. E. Lundvall (Baltimore Gas and Electric Company) from D. Eisenhut (NRC), Docket No. 50-317, September 15, 1983.
568. Memorandum for W. Houston and L. Rubenstein from F. Schroeder, '! Request for Reactor Systems Branch and Auxiliary Systems Branch Support for Plant Visits on USI A-45," November 28, 1983.
901. Memorandum for T. Combs from H. Denton, " Revised SRP Section 9.2.1 and SRP Section 9.2.2 of NUREG-0800," June 24, 1986.
902. Memorandum from J. Sniezek and R. Fraley from H. Denton, " Resolution of Generic Issue No. 36, ' Loss of Service Water,'" May 13, 1986.
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ISSUE 74: REACTOR COOLANT ACTIVITY LIMITS FOR OPERATING REACTORS DESCRIPTION Historical Background This issue was raised 5" by DSI in June 1983 and addresses the concern that-several operating PWRs and BWRs either lack iodine coolant activity LCOs or have inadequate LCOs such that accidents involving the release of coolant cannot be satisfactorily precluded from causing offsite doses in excess of 10 CFR Part 100 guidelines.
Prior to 1974, limiting conditions of operation (LCO) for coolant activity were determined on a plant-by plant basis, postulating accidents such that subsequent releases and exposures were an appropriately small fraction of the 10 CFR 100 guidelines. Gross activity limits were typically specified with an assumed isotope spectrum. The limiting accident was represented by the radiological consequences of a postulated SGTR in a PWR or a steam line break in a BWR.
Similarly, the allowable secondary activity in a PWR was limited by a postu-lated secondary coolant steam line break. Many plants licensed before 1974 only sample for gross activity and do not identify specific iodine isotope concentrations.
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\V I In May 1974, standard technical specificaticas (STS) were proposed for limiting the dose equivalent Iodine-131 coolant activity concentrations. The purpose was to establish uniform concentration limits for all plants. The basis for this uniform STS LC0 is that the calculated exposure resulting from SGTR or steam line break accidents be below the 10 CFR 100 guidelines at the site having the worst meteorological characteristics. The STS limiting equilibrium 1-131 concentrations are 10 8 and 10 7 Ci/gm of water in the primary and secondary coolant, respectively, for PWRs, and 2 x 10 7 Ci/gm of water for BWRs.
All plants licensed since December 1974 have-implemented the reactor coolant activity level STS. In addition, some plants licensed prior to 1974 have adopted the reactor coolant activity level STS. A review of the licensed plants in 1982 and a recent updating indicates that all plants have limits on coolant activity, but 10 PWRs have no LC0 on reactor coolant iodine activicy concentra-tion and one PWR and twenty BWRs have LCOs that are higher t h the STS LCO.
Since this generic issue was established, NRR, through the Steam Generator Tube Rupture Task Force (see Issue 66), has recommended that the STS reactor coolant activity level LC0 be backfit to all remaining PWRs that have not yet adopted them. Because of considerations introduced by the presence of steam generators, it is appropriate to consider PWRs separately. Furthermore, this issue does not include any information or analysis on PWRs not included in Issue 66.
Therefore, the PWR aspects of this generic issue are subsumed in Issue 66 and
] Issue 74 is reduced to consideration of backfitting the STS LC0 for reactor coolant activity level for only those 20 BWRs that have not yet adopted them.
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Revision 1 l Safety Significance The calculated radiological consequences for some accidents are dependent upon the magnitude of the reactor coolant iodine equivalent activity assumed in the dose calculation model. However, the magnitude of reactor coolant activity levels will have a significant effect for only non core-melt accident conse-quences. Non-core-melt accidents are not major contributors to the expected public risk from nuclear power plant operation.
In addition to standard monitoring, sampling, and reporting requirements, STS promote uniform characteristics in shielding, personnel protection, and coolant cleanup system capacity.
PRIORITY DETERMINATION A technical analysis of the proposed resolution for this issue was performed by PNL and is documented in Supplement 4 to NUREG/CR-2800.64 As stated above, resolution of this issue is applicable to only the 20 operating BWRs that have not yet adopted the STS requirements for reactor coolant activity. These plants are assumed to have an average remaining lifetime of about 25 years.
Frequency / Consequence Estimate Currently, BWRs operate with average I-131 concentrations of approximately 10 8 Ci/gm of coolant which is a factor of 20 below the STS LC0. Careful management of coolant activities, based primarily on the desire to control ORE, has resulted in these low levels being observed. Implementation of the STS LCOs at the 20 affected BWRs is, therefore, not expected to result in lower observed average equilibrium iodine concentrations. As a result, resolution of this issue is not expected to result in a public risk reduction for non-core-melt accidents which are assumed to occur when the reactor coolant activity is at the equilibrium condition. It should be noted that implementation of the STS LCOs would impose an increase in reactor coolant surveillance requirements and would, therefore, reduce the uncertainty in observed concentration levels.
However, since observed average concentration levels are a factor of 20 lower than the STS LCO limit, our conclusion that resolution of this issue would not be expected to result in a public risk reduction would not be altered by a reduced uncertainty in observed concentration levels.
Situations can be postulated where a plant could operate with iodine concentra-tions above the STS LC0 due to the iodine spiking phenomenon. This situation has been addressed in the prioritization of Issue B-65, " Iodine Spiking." Reso-lution of Issue B-65 was also assumed to result in the imposition of new reactor coolant activity LCOs which would be derived from a better understanding of the iodine spiking phenomenon. It was assumed that the new LCOs which might be imposed would not allow for greater iodine concentration levels than those allowed by the STS LCOs. The public risk reduction thus afforded was estimated for both PWR and BWR plants. The public risk reduction estimated for BWR plants by limiting iodine spiking peak concentrations was less than 0.1 man-rem, again because the observed average activity levels in BWR plants are significantly lower than the STS LC0 activity levels.
We, therefore, conclude that resolution of this issue, i.e., backfitting the STS LCOs to the 20 BWRs that have not yet adopted them, would not be expected to result in a public risk reduction.
06/30/86 3.74-2 NUREG-0933
Revision 1 Cost Estimate v
Industry Cost: Resolution and implementation of this issue were assumed-to place the following requirements on the 20 affected BWRs. All the plants were assumed to process a TS change at a cost of one man-month of licensee staff ef-fort per plant. The STS LCOs for BWR plants require sampling for dose equiva-lent iodine once every 31 days and gross activity sampling at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Implementation of the STS LCOs would require additional sampling and isotopic analysis capability at some plants. Five of the affected BWRs were assumed to require additional sampling and analysis equipment estimated at
$250,000 plus one man month for installation and operational verification. At an assumed cost of $2,270/ man-week, an industry cost for implementation of
$1.54M is estimated. Imposition of the STS LC0 is estimated to require about 44 additional samplings and analyses per year at each of the 20 affected BWRs.
At 2 man-hours sample, the present worth of the total additional industry operating cost is estimated to be $2.5M. Adding the industry implementation and operating costs, a total industry cost of $4M is estimated.
NRC Cost: NRC impact is limited to the effort required.to issue an order to the 20 affected BWRs to implement the STS LCO and the staff effort to audit the licensee TS changes. The estimated NRC cost ($25,000) is negligible in com-parison to the estimated total industry cost ($4M).
Value/ Impact Assessment '
Since there is no perceived public risk reduction for the resolution of this g issue, the value/ impact score is 0 man-rem /$M.
v Other Considerations Resolution of this issue is assumed to require the installation of additional sampling and analysis at a quarter of the 20 affected BWRs and the gathering and analysis of about 44 additional reactor coolant samples per year at all of the affected plants. Since these activities are expected to be performed in a low level radiation field, additional ORE is anticipated. Assuming a 25 millirem /hr field, we calculated 5 man rem ORE for installation of additional equipment and 1100 man-rem ORE for the gathering and analysis of additional reactor coolant samples over the remaining lifetime of the 20 affected plants.
On the other hand, imposition of STS LCOs for reactor equilibrium coolant activ-ity levels would provide a means to limit reactor coolant activity levels during those infrequent instances in which severe fuel leaks develop. The limiting of reactor coolant activity levels for those instances of operation with " bad" fuel would reduce activity levels in the vicin ty of the reactor coolant system and, i
therefore, would be expected to result in a reduction of ORE to the plant operating staff.
We have made a probabilistic estimate of the expected savings in ORE to plant personnel for the 20 affected BWRs over the next 25 years by backfitting STS coolant activity limits. In the analysis we have assumed that the average BWR coolant activity is 0.01 pCi/gm I-131 equivalent and that coolant activity levels i could exceed the STS LC0 (0.2 pCi/gm) by about a factor of 3 before other con-
. trols such as steam line activity level or plant stack activity level would require corrective action. Examination and evaluation of historical data on 06/30/86 3.74-3 NUREG-0933
Revision 1 operator exposure at BWR plants lead us to the conclusion that about 20% of the annual exposure of plant personnel can be directly affected by reactor coolant activity levels, i.e., about 180 man-rem /RY of the average annual exposure of 900 man-rem /RY may be due to the current average reactor coolant activity level of 0.01 pCi/gm. When combined with a historical background which indicates somewhere between 2 to 6 instances of major BWR fuel leakage, we estimate that imposition of the BWR STS limit of 0.2 pCi/gm I-131 equivalent at the 20 affected plants could result in a total reduction in ORE of between 600 to 1700 man-rem over the remaining lifetime of the 20 plants. We also calculated the potential reduction in ORE from plant cleanup for mitigated LOCAs by limiting the reactor coolant I-131 activity level to no greater than 0.2 pCi/gm and estimated an average total ORE savings of 125 man-rem.
When the above increase and reductions in ORE are summed, we arrive at a con-clusion that imposition of the BWR STS requirements on reactor coolant activity could result in a range of ORE change from a 375 man-rem increase to about a 725 man-rem reduction over the lifetime of the 20 affected BWRs.
CONCLUSION The resolution and implementation of this generic issue is not expected to result in any appreciable offsite (public) risk reduction, but can result in additional costs for the licensees of the 20 affected BWRs. Estimates of ORE indicate an increase for additional coolant sampling, a very small projected averted ORE due to plant cleanup in the event of mitigated LOCAs with " bad" fuel, and, at best, a small decreasa in ORE due to operation of the plant with lower peak coolant activity limits. The overall effect of resolution and implementation of this issue could range from a small ORE decrease to a small ORE increase which could very well negate each other. Thus, we do not view ORE as a signi-ficant consideration and recommend that the issue be DROPPED.
REFERENCES
- 64. NUREG/CR-2800, " Guidelines for Nuclear Power Plant Safety Issue Prioritiza-tion Information Development," U. S. Nuclear Regulatory Commission.
519. Memorandum for W. Minners from L. Hulman, " Generic Issue on Iodine Coolant Activity Limitng Conditions for Operation," June 10, 1983.
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Revision 1 V
ISSUE 99: RCS/RHR SUCTION LINE VALVE INTERLOCK ON PWRs DESCRIPTION Historical
Background:
On April 17, 1984, a DSI memorandum 788 on the subject of RHR interlocks for W plants described staff concerns that the design basis for RHR interlocks had been misunderstood and that these concerns had not been adequately pursued in
.recent reviews. As a result, DST was requested to prioritize this concern as a generic safety issue.787 Interlocks are provided to assure that there is a double barrier (two closed valves) between the RCS and RHR systems when a plant is at normal operating conditions, i.e., pressurized and not in the RHR cooling mode. A related issue (Issue 96) addresses the concern of assuring that both series RHR isolation-valves are closed during normal power operation. Issue 99 is concerned with the inadvertent closing of these valves when the RHR system is in use.
Two basic features are incorporated in the interlock design: (1) an automatic closure signal on high RCS pressure (typically 600 psig), and (2) a block of the manual open signal at a lower RCS pressure (typically 425 psig). The auto-O)
(' closure setpoint is generally set higher than the design pressure of the RHR system. However, overpressure protection of the RHR system during RHR cooling is provided by relief valves and not by the slow-acting RHR suction valves. The block setpoint is lower than the RHR system design pressure to preclude opening of either RHR suction valve when the RCS is at a higher pressure.
In the W design, 2 interlock channels are provided such that 1 channel is used to interlock the operation of one RHR suction valve and the other channel is used for the other valve. The same interlock configuration is used in W plants for designs that have 1 or 2 RHR drop lines from the RCS. When either channel is in a tripped state, its associated suction valve will automatically close if it is open. Since the relays used for this interlock are deenergized to initi-ate valve closure, a loss of the instrument bus used for either channel will re-
) sult in a loss of RHR cooling due to inadvertent closure of one of the suction valves.
Safety Significance The loss of one instrument bus or disablement of one logic channel will result in the automatic closure of one of the RHR suction line isolation valves. In the RHR cooling mode, such closure gives rise to the potential for RHR pump damage and loss of decay heat removal by the RHR system. This safety concern applies to all W reactors.
Possible Solutions O
h The proposed resolution to this issue that was assumed for cost estimation pur-poses consists of the following parts:
06/30/86 3.99-1 NUREG-0933
1 1
R:visicn 1 (1) Review and document the design basis for the RHR suction valve interlock.
(2) Develop interim operating procedures until changes to the logic and control for the RHR system can be implemented.
(3) Change the logic configuration that controls the valves from a one-of-one configuration to a two-of-two configuration. Improvements in detecting and alarming of the loss of RHR coolant flow would be made.
(4) Changes to the plants' TS.
PRIORITY DETERMINATION Frequency Estimate NSAC-52798 lists 27 events through 1981 that involved loss of RHR flow due to suction valve closure. Two of these events occurred as the result of a pressure rise in the primary system. The other 25 events resulted from causes other than an actual pressure rise and occurred during 206 RY of operating experience at PWRs. This experience results in a frequency of 0.12 unplanned RHR suction valve closures per plant year. Of these 25 closures, 22 events involved the closure of only 1 valve and 3 events resulted in the closure of both valves.
Thus, 88% of the reported events were independent channel failure events and 12% can be potentially classified as common-cause related.
When in the refueling mode and the water level is 23 feet above the core, only one RHR train must be operable. Closing the suction valve could cause cavita-tion and damage to the pump and leave no RHR train operable. However, it would take many hours for the level to boil down and uncover the core. RHR cooling could be restored in a few hours. In addition, the fuel pool cooling system could be used. Therefore, this case would have a small associated risk.
In all other modes, two RHR trains are required to be operable while only one is usually operating. If the RHR valves close causing cavitation and damage to the operating RHR pump, the other RHR train would still be operable. The NSAC data 798 show that the operator successfully reopened the inadvertently closed valves immediately in all but one event. In this event at Davis-Besse, it took 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to restore RHR cooling because of the need to refill and vent the system. Yet, in this lengthy delay, no sustained damage occurred to the system components. However, due to the long time interval involved before re-storing RHR cooling, this event was counted as an RHR system failure. Thus, an unavailability of 0.04 is assumed but believed to be overly conservative. If the valve cannot be reopened, either the steam generators or the charging pumps could be used as alternate means of cooling.
Based upon engineering judgment, the unavailability of the main and/or auxiliary feedwater during RHR operation is estimated to be 0.1 and the unavailability of the charging pumps is estimated to be 0.01. These may be overly optimistic since there is no TS requirement for the availability of these systems in the cold shutdown modes. Further, maintenance and testing is often performed on these systems during the RHR cooling modes. Thus, the unavailability of core cooling is estimated to be the product of (0.12 event /RY)(0.04)(0.1)(0.01) or 4.8 x 10 8/RY which, assuming no further actions are taken, becomes the expected 06/30/86 3.99-2 NUREG-0933
Rtvision 1
/ core-melt frequency resulting from an inadvertent closure of one or both RHR h suction valves.
Changing the logic system from a one-of-one system to a two-of-two system will reduce the independent failure frequency contribution of one valve closure from 0.12/RY to 0.003/RY. With the common mode contribution remaining the same (0.015/RY), the revised frequency of incorrect valve closures reduces to 0.018/RY by the revised logic configuration. Improved procedures and alarms are assumed to reduce the human error of failing to reopen the RHR isolation valves from 0.04 to 0.02 per event. The changes in failure rates reduce the expected core-melt frequency from an RHR valve being closed to (0.018)(0.02)(1)(0.01)/RY or approximately 3.6 x 10 7/RY. This represents a core-melt frequency reduction of 4.4 x 10 8/RY.
Consequence Estimate The expected radiological consequences from this issue are expressed in whole body tran-rem dose based upon the radioactive release categories described in WASH-1400.18 The computer program CRAC284 applied to a typical midwest site meteorology (Braidwood) was used for the dose calculation. An average popula-tion density of 340 persons per square mile was used over an area which extended from an exclusion zone of one-half mile about the reactor out to a 50-mile radius about the reactor.
A core-melt resulting from the loss of the RHR system would result in an ac-p p)
V cident similar to the T MLUi sequence described in the Oconee RSSMAP analy-sis.54 The release, given a core-melt, occurs in the following categories with-the respective probability and dose:
Category Probability Dose (man-rem) 3 0.5 5.4 x 108 5 0.0073 1.0 x 108 7 0.5 2.3 x 103 A core-melt frequency reduction of 4.4 x 10 8/RY results in a dose reduction of 12 man-rem /RY. For the 30 existing reactors with an average remaining life of 27.7 years and 28 new plants with an expected life of 30 years, the total risk reduction for this issue amounts to 20,000 man-rem.
Cost Estimate
- Industry Cost
- The cost estimate addresses the four actions proposed as the resolution of this issue. The review and documentation of the design basis of the RHR suction valve interlocks is expected to require 4 man-weeks which, at a rate of $2,270/ man-week, results in a cost of $9,080/ plant. The development of interim operating procedures and operator training is estimated to total 5 man-weeks / plant or $11,350/ plant. Hardware costs to modify the logic system and install the RHR flow alarms are estimated to be $4,000. An additional 6 man-weeks ($13,620) will be required for engineering and installation costs. The total hardware modification cost is estimated to be $17,600. TS changes are estimated to take 4 man-weeks or $9,080. Thus, the costs for issue resolution O( are estimated to be $47,200/ plant. Plants not having an operating license are expected to have a lesser cost but, due to the advanced stages of construction, 06/30/86 3.99-3 NUREG-0933
Revision 1 the reduction is not expected to be significantly less. Modifications to plant hardware are expected to be performed during a refueling outage and would obvi-ate the need to include replacement fuel costs. No significant additional main-tenance costs over the currently existing configuration are envisioned. Thus, for all 58 plants, the total industry cost is estimated to be $2.7M.
NRC Cost: It is estimated that NRC costs associated with the issue resol,ution can be accommodated in a total of 8 man-weeks or $38,000.
Thus, the total cost associated with the resolution of this issue is
$(2.7 + 0.038)M or approximately $2.74M.
Value/ Impact Assessment Based on an estimated public risk reduction of 20,000 man-rem, the value/ impact score is given by:
3 = 20,000 man-rem
$2.74M
= 8,000 man-rem /$M Other Considerations The analysis did not consider the possible increase in the chance of an inter-facing systems LOCA which might result because the logic changes reduced the reliability of the interlock function. It is presumed that the reliability of a one-out-of-one logic is the same as a two-out-of-two logic.
The ORE is estimated to be 2.25 man-rem / plant for work involved with hardware modifications inside the containment. This would result in a total worker dose of 176 man-rem. The accident avoidance occupational dose reduction is estimated to be 146 man-rem.
The industry cost savings due to accident avoidance are estimated to be (4.4 x 10 6 accident /RY)($1.65 billion / accident) or $7,260/RY.
Consideration also should be given to those cost savings which result from the prevention of incidents producing long interval RHR inoperability, but do not result in damage to the core. Such incidents may result in plant shutdown longer than anticipated to investigate the causes of the inoperability and to assure the adequate corrective actions have been taken. Assuming that the out-age extension lasts 2 weeks, the replacement power costs (estimated at $500,000/
day) are $7M. At the current frequency of long interval outage events, the savings per plant resulting from incident avoidance are $90,000.
CONCLUSION Based on the averted public risk and the value/ impact score, this issue has a HIGH priority ranking. The public risk may be underestimated if the feedwater and injection alternatives are not as available as predicted. Additional recommendations made in AE00/C5039 9 following the prioritization of this issue will be addressed in the resolution of the issue (References 910 and 911).
3.99-4 NUREG-0933 06/30/86
Revision 1 l REFERENCES
- 16. WASH-1400 (NUREG-75/014), " Reactor Safe'ty Study, An Assessment of Accident. Risks in U.S. Commercial Nuclear Power Plants," U.S. Nuclear Regulatory Commission, October 1975.
- 54. NUREG/CR-1659, " Reactor Safety' Study Methodology Application Program,"
U.S. Nuclear Regulatory Commission, 1981.
- 64. NUREG/CR-2800, " Guidelines for Nuclear Power Plant Safety Issue i
Prioritization Information Development," U.S. Nuclear Regulatory Commission.
796. Memorandum for R. Mattson from T. Dunning "RHR Interlocks for Westinghouse Plants," April 17, 1984.
t 797. Memorandum for F. Rowsome from W. Houston, "RCS/RHR Suction Line Valve Interlock on PWRs," August 27, 1984.
f 798. NSAC-52, " Residual Heat Removal Experience Review and Safety Analysis; Pressurized Water Reactor," Nuclear Safety Analysis Center, January 1983.
909. AE0D/C503, " Decay Heat Removal Problems at U.S. Pressurized Water Reactors," Office for Analysis and Evaluation of Operational Data, U.S.
i Nuclear Regulatory Commission, December 1985.
910. Memorandum for H. Denton from C. Heltemes, " Case Study Report - Decay Heat Removal Problems at U.S. Pressurized Water Reactors,"
i December 23, 1985.
911. Memorandum for C. Heltemes from H. Denton, "AEOD's Report on Decay Heat Removal Problems at U.S. PWRs," February 10, 1986.
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U ISSUE 114: SEISMIC-INDUCED RELAY CHATTER DESCRIPTION i
Historical Background This issue was identified in a RRAB memorandum 815 in March 1985 and concerns the possibility of relay contact chatter during a seismic event and its resulting effect upon safety and safety-related electrical control systems and their abilities to provide for and maintain a safe plant shutdown.
Various regulatory requirements address seismic design requirements and defini-tions. These include, but are not limited to: 10 CFR Part 100,197 Appendix A,
" Seismic and Geologic Siting Criteria for Nuclear Power Plants"; SRPti Sec-tions 2.5.1, 2.5.2, and 3.10; Regulatory Guide 1.29,816 " Seismic Design Clas-sification"; and Regulatory Guide 1.100,817 " Seismic Qualification of Electric Equipment for Nuclear Power Plants." Currently, NRC requirements require a nuclear plant to withstand the. effects of an SSE and to assure: (1) the integ-rity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and to maintain it in a safe condition, or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposure. The SSE is defined as that earthquake which is based upon an evaluation of the maximum earthquake potential considering the i
(O) regional and local geological and seismological and specific characteristics of local subsurface material. It is that earthquake which produces the maximum vibratory ground motion for which certain structures, systems and components are designed to remain functional.
Of concern is the capability of relays to perform at the SSE. For earthquakes which exceed the SSE (a low probability event), the ability to assure a margin j for relay chatter is sought.
The RRAB memorandum 815 identified a number of activities which address aspects
- of this issue. These include
- (1) the NRC contract with Future Resource Asso-ciates, Inc., which is investigating whether seismic PRAs can be improved by detailed analyses of equipment failures and operator-responses; (2) Seismic ~
Design Margins Program which is to~ determine the seismic margins in the existing l plants in the eastern United States; and (3) USI A-46 which addresses the seismic
- qualification of equipment in operating plants. As part of USI A-46, there are efforts for the evaluation of the qualification _ test adequacy of relays required
- to be functional to assure safe plant shutdown and for the collection of fragi-lity data by the Seismic Qualification Utility Group. Currently, there is a review group within NRC, supported by a corresponding industry group, working on the criteria for determination of relays that are needed to perform during a seismic event. The NRC review group and the USI A-46 task manager will discuss the need to include safety and safety-related electrical components as part of i
the industry program during the scheduled June 1986 meeting.
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Safety Significance A large number of relays are used to control many devices in nuclear power plants such as pumps, valves, and circuit breakers. In addition, much of the control logic for system initiation and control is accomplished with relays.
Relay control may be accomplished by relay actuation either in the energized or the deenergized state.
Seismic-induced relay chatter is the opening and/or closing of relay contacts due to the influence of seismic accelerations either directly or indirectly.
It is possible that this phenomena may result in important equipment being rendered inoperable as well as the loss of systems status indications. Valves may be erroneously opened or closed, pumps may be shut off, diesel generators may be rendered inoperable, etc. Operator actions, not only in the control room but also at various locations in the plant, may be required to realign relays and control circuits to restore required systems to an operable con-figuration and maintain core cooling. Reactor trip is very likely to result from causes other than relay chatter.
PRIORITY DETERMINATION Quantification of the risks resulting from seismic-induced core-melt accidents involves a large amount of uncertainty. These uncertainties include a large variation in the estimate of recurrence frequency of seismic events and also in the predicted seismic fragility. The majority of the fragility prediction data employed in performing PRAs is based upon expert judgment; very little is based upon actual test results. Not only is there a large uncertainty associated with fragility data but also the peak amplitude of motion experienced by a component may vary with changes in the frequency character of the seismic input resulting from the frequency response of the mounted structure such as cabinets, panels, etc. NUREG/CP-0070 9 ts provides insights into the many uncertainties associated with performing PRA-type analyses of seismic events.
In addition to the previously described uncertainties, the response of relays to seismic acceleration has large variations depending upon the relay design (type, size, structure), the state (energized or denergized), the frequency spectra of the input motion, and the magnitude of the input motion. It has been found that relay response is non-linear, i.e., the fragility does not necessarily decrease with increases in input frequency or acceleration.
Further, the effect of relay chatter is very dependent upon the dynamic re-sponse characteristics of the circuit in which the relay contacts are employed.
Circuits with fast time constants or response may be very sensitive to relay chatter while circuits with slow response characteristics may not be perturbed by relay chatter.
USI A-46 will provide insights into the effects of relay chatter from seismic events. It will identify systems whose performance may be degraded by relay chatter. Further, it will determine the effect of relay chatter upon these required systems and develop the procedural guidelines, including human responses, to mitigate these effects.989 Most important, the results of USI A-46 will assure that the systems required to bring the plant to a safe shutdown are not disabled by relay chatter.
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g CONCLUSION This issue is currently covered in USI A-46 for relay performance at the SSE level. .For-greater than SSE events, the Seismic Margins Program and the related programs to collect fragility data address relay capability. Thus, there is no need for an additional, separate issue at this time.
REFERENCES
- 11. NUREG-0800, " Standard Review Plan," U.S. Nuclear Regulatory Commission.
197 Code of Federal Regulations, Title 10, Energy.
, 915. Memorandum for W. Minners from A. Thadani, " Seismic Included Relay Chatter l Issue," March 22, 1985.
916. Regulatory Guide 1.29, " Seismic Design Classification," U.S. Nuclear l
t Pegulatory Commission, (Rev. 3) September 1978.
917. Regulatory Guide 1.100, " Seismic Qualification of Electric Equipment for Nuclear Power Plants," U.S. Nuclear Regulatory Commission, (Rev. 1)
August 1977.
918. NUREG/CP-0070, " Proceedings of the Workshop on Seismic and Dyanamic Fragi-j lity of Nuclear Power Plant Components," U.S. Nuclear Regulatory Commis-l ston, August 1985.
l s ,, 919. NUREG-10:0, " Seismic Qualification of Equipment in Operating Nuclear Power Plants," U.S. Nuclear Regulatory Commission, (Draf t) August 1985.
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U J
i ISSUE 122: DAVIS-BESSE LOSS OF ALL FEEDWATER EVENT OF JUNE 9, 1985 -
SHORT-TERM ACTIONS The loss of all feedwater event at Davis-Besse on June 9, 1985 resulted in the formation of an NRC project team to investigate the event. The team's findings were published in NUREG-1154888 and were subsequently reviewed by DL. As a result of DL's review, the following items were identified as candidates for short-term staff actionsas and were forwarded to DST for prioritization:ss7
- 1. Potential inability to remove reactor decay heat due to questionable reliability of the auxiliary feedwater system caused by any or all of the following:
- a. Loss of all auxiliary feedwater due to common-mode failure of AFW pump discharge isolation valves in closed position.
- b. Excessive delay in recovery of auxiliary feedwater due to difficulty in restarting AFW pump steam driven turbines, if turbines are tripped.
- c. Interruption of auxiliary feedwater flow due to failures in steam and feed line break accident mitigation features b (e.g., SFRCS).
- 2. Adequacy of emergency procedures, operator training and available plant monitoring systems for determining need to initiate feed--
and-bleed cooling following loss of steam generator heat sink.
- 3. Physical security system constraints which could deny timely opera-tor access to vital equipment and inhibit operator from performing local manual operatiors called for in emergency procedures.
The above items formed the basis for Issue 122 but were prioritized separately as shown below. The identification of each item prioritized follows the num-bering system established in the DL memorandum. ass The prioritization results
- are summarized in Table 3.122-1.
ITEM 122.1: POTENTIAL INABILITY TO REMOVE REACTOR DECAY HEAT
- During the loss of main feedwater event, the reactor scrammed and the AFW system should have actuated and supplied feedwater to the steam generators to i
enable them to remove decay heat. However, in this process several failures occurred, three of which are of significance here.
(1) An operator attempted to start the two AFW trains manually, in addition to the automatic signal on low steam generator water level. Unfortunately, f the operator pressed the wrong buttons, sending erroneous " low steam generator pressure" signals to both AFW trains. The AFW control systems 06/30/86 3.122-1 NUREG-0933
TABLE 3.122-1 Item Staff Action Priority 1.a Common Mode Failure of HIGH AFW Pump Discharge Isolation Valves in Closed Position 1.b Excessive Delay in MEDIUM Recovery of Auxiliary Feedwater 1.c Interruption of Auxiliary HIGH Feedwater Flow
- 2. Adequacy of Emergency HIGH Procedures, Operator Training and Available Plant Monitoring Systems
- 3. Physical Security System LOW Constraints then caused both AFW isolation valves to close. Thus, neither steam generator could receive any water. In essence, the operator caused a common mode failure.
(2) Both AFW turbines tripped on overspeed. The overspeed trips on such turbines usually have to be reset at the turbine, not from the control room.
(3) In attempting to recover the AFW system, the operators reset the erroneous signals. However, the AFW isolation valves did not open. In spite of several attempts, the plant operators were unable to open these valves from the control room, and ultimately had to open them by hand.
The three parts of this item are evaluated separately below.
ITEM 122.1.A: FAILURE OF ISOLATION VALVES IN CLOSED POSITION DESCRIPTION Historical Background This item addresses Findings 4, 5, 6, and 15 in Section 5.2.5 of NUREG-1154.88c, The particular issue deals with a potential inability to remove reactor decay heat because of loss of all auxiliary feedwater due to the third common mode failure discussed above. This is the failure of AFW pump discharge isolation valves to reopen on command after they had closed.
06/30/86 3.122-2 NUREG-0933
Safety Significance With the main feedwater out of service (the transient initiator), a spurious closing of these AFW valves cannot easily be rectified, leaving only feed-and-bleed techniques available for removal of decay heat. Westinghouse FWRs gen-erally do not have such motor operated isolation valves in the AFW discharge lines, but some W plants plus roughly 16 plants designed by B&W and CE in addition to Davis-Besse may be susceptible to this problem.
Possible Solutions The failure of the Davis-Besse AFW valves to reopen was ultimately traced to the torque limit a'nd bypass switches which control the motor operators of the valves. In essence, the high differential pressure across the closed valves necessitated a relatively large force for valve motion. The motor control switches were not adjusted to accommodate such a force. Such a failure can happen in two ways. First, the switches can be inadvertently mis-adjusted during routine maintenance. Second, the valve may be correctly maintained but the actuation system is not designed to provide for an open command to these valves (in some PWRs), or the torque necessary to reopen these valves under some conditions may be beyond the design capacity of the valve actuators. In the case of Davis-Besse, the valves were designed to close (which is their intended safety function), but apparently less attention was paid to their ability to reopen.
The solutions are implicit in the causes. For this prioritization we assume Oi that the actuation system is equipped to issue open commands so the solution is to verify that the valves, as designed, are capable of reopening in the presence of a differential pressure, and upgrade the calibration and maintenance procedures. e PRIORITY DETERMINATION Frequency Estimate To estimate accident frequencies, we will follow the examplesas in which the relatively simple transient classifications of the Oconee RSSMAP study s4 were used, but frequency and probability estimates were taken from the more modern sources such as the more detailed PRA of Oconee 3 done by EPRI and Duke Power Co.889 The affected sequences in the R$5 MAP studyS4 are T M(LOPNRE)LU, t T 2MLU and Tdi(PCSNR)LU, where Ti is a loss of offsite power (LOOP) transient with an assumed frequency of 0.05 transient /RY (or more).890 T2 is a non-recoverable loss of the Power Conversion System caused by other than a LOOP, with an assumed frequency of 0.64 transient /RY based on the Oconee PRA.889 p' T3 is a transient with the Power Conversion System initially d available, with an assumed frequency of 5.7 transients /RY also based on the Oconee PRA.sso 06/30/86 3.122-3 NUREG-0933
M This is a failure of the power conversion system. The probability is unity for Ti and T 2 sequences. For Ta sequences, we will use 3.7 x 10 3, obtained by summing the failure modes listed in Section A8.3.8 of the Oconee PRA.889 LOPNRE This is the probability of non-recovery of offsite power in 40 minutes after a LOOP event. We estimate this to be roughly C.25, based on the generic curves given in NUREG-1032.890 PCSNR This is the probability of non recovery of the Power Conversion System (really, main feedwater) in 30 minutes.
The Oconee PRA889 uses 0.3 for a similar event (event REFDW2). It must be remembered that this figure is somewhat optimistic because of the ability to cross-connect at the Oconee site.
L is failure of the AFW system.
U is a failure to cool the core via feed-and-bleed. For Oconee and most other plants, this is essentially a failure of the high pressure ECCS. The assumed probability is 0.015 based on the Oconee PRA 889 The unquantified parameter is AL, the change in the AFW failure probability to be attributed to this issue. It is composed of three factors: the probability of spurious isolation, the probability of failure to reopen on demand, and the probability of failure of reopening (in time to prevent core damage) by manual action.
Davis-Besse has been in operation for eight years. The licensee reports a frequency of loss-of-feedwater events of 0.67/ year.891 Thus, the AFW system has had about five real challenges. One of these was the June 9, 1985 event where an operator inadvertently pushed the wrong button and caused a spurious isolation. One would therefore expect the spurious isolation rate to be roughly one in five AFW demands, or 20%, and dominated by human error. How-ever, it would be naive to assume that this event (and its associated extended shutdown) has gone unnoticed in the control rooms of other plants. Nor can it be assumed that all other plants have an AFW control panel like that of Davis-Besse. On the other hand, the AFW discharge isolation valves may be initially closed at the time of the demand, as they were at the outset of the accident at
, TMI-2. We will assume a 5% minimum likelihood of sourious or inadvertent AFW l isolation and assume further that plants with a high (e.g., 20%) likelihood I will be addressed by Iten 122.1.C.
Next is the question of failure of the isolation valves to open on demand.
As was mentioned before, this can happen either by errors in' maintenance or by a lack of foresight in design. For the case of errors in' maintenance, we
- turn to the valve failure data tabulated in NUREG/CR-2770.892 Of the 393 MOV l failures listed, 75 involved torque limit or bypass switches, and 34 of these l (about 8.7% of all the failures) appeared to be adjustment or calibration
! errors. Since the same crews and procedures are used on all AFW trains, these i
failures are very likely to be present on all trains. Given a failure on one train, we will assume an 8.7% probability 892 that the failure was due to im-proper torque or limit switch adjustment and that the analogous valves ,n the 06/30/86 3.12P4 NUREG-0933 1
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r m
(v) redundant trains will also fail.
rate of 6.4 x 10-3/ demand.
The RSSMAP studys4 used an M0V control failure The probability of failure to reopen due to main-tenance error is the product of these two figures,.or 5.6 x 10 4 For the case of lack of foresight in design, there is no extensive tabular data. This particular scenario, by its very nature, will affect both valves.
However, this does not mean that both valves necessarily will fail to open.
NUREG-1154sse describes tests of the actual valves at Davis-Besse, five of which were at a full differential pressure of 1050 psid. One valve failed to open twice. The other valve failed onca but opened successfully two times.
Thus, for a two-train AFW system, the probability of neither valve opening would be expected to be on the order of (1 x 0.33), or 33%, based on this admittedly sparse data.
Finally, the probability of the operator failing to reopen the valves manually must be estimated. In the case of the Davis-Besse event the spurious closure occurred about six minutes into the event. NUREG-1154885 mentions a 30 minute interval before core damage would be expected. Thus, the operators had about 24 minutes in which to reopen at least one valve. In actual fact, it took an average of 7.5 minutes (about a third of the available time) to open these two valves. This is plenty of margin and would normally imply a failure rate (due to timeout) of a percent or two. However, it should be noted that, except for one button pushing error (which is understandable in the light of hindsight),
this operating crew performed very well. The shift supervisor and his assistant were astute in diagnosing the AFWS misalignment (while being faced with a bar-(7 i
}
rage of other information) and took.the correct action to manually open the auxiliary feedwater block valves. We will assign a 10% probability of failure V
to manually reopen the valves, based purely on judgment of the human factors aspects.
Putting these factors together, the AFW failure probability is the product of a 5% probability of inadvertent AFW isolation, a 33% probability that neither valve will reopen on demand, and a 10% probability that manual opening will'not be attempted or will fail to be accomplished in time. The product is 1.7 x 10-3/ demand. In addition, no solution is perfect. We will assume that any resolutions adopted will be at least 90% effective. Thus, the change in AFW failure probability will be on the order of 1.5 x 10-3 The change in core-melt frequencies can now be estimated. The cut sets are:
T iM*LOPNRE*AL*U 3 x 10 7/RY T2M*AL*U 1.5 x 10 5/RY Ta*M*PCSNR*AL*U 1.5 x 10 7/RY Total AF 1.5 x 10 5/RY Under the assumption that one plant will find and correct the problem, the <
core-melt frequency is 1.5 x 10 5/ year. '
Consequence Estimate 7
Normally, accident sequences such as the ones discussed in the previous section would be distributed across a spectrum of containment failure modes 06/30/86 3.122-5 NUREG-0933
in a variety of ways. However, because the sequences of interest here are similar in their final stages prior to core-melt, all three sequences will be distributed across the containment failure modes in the same manner.
All three principal accident sequences involve a core-melt with no large breaks initially in the reactor coolant pressure boundary. The reactor is likely to be at high pressure (until the core melts through the lower vessel head) with a steady discharge of steam and gases through the PORY. These are conditions likely to produce significant hydrogen generation and combustion.
The Zion and Indian Point PRA studies used a 3% probability of containment fail-ure due to hydrogen burn (the " gamma" failure). We will follow this example and use 3%, remembering that specific containment designs may differ significantly from this figure. In addition, the containment can fail to isolate (the " beta" failure). Here, the Oconee PRA889 figure of 0.0053 will be used. If the con-tainment does not fail by isolation failure or hydrogen burn, it will be assumed to fail by base mat melt-through (the " epsilon" failure).
Using the usual prioritization assumptions of a central midwest plains meteo-rology, a uniform population density of 340 persons per square mile, a 50-mile radius and no ingestion pathways, the consequences are:
Failure Percent Release Consequences Mode Probability Category (man-rem) gamma 3% PWR-2 4.8 x 106 beta 0.5% PWR-5 1.0 x 106 epsilon 96.5% PWR-7 2.3 x 10 3 The " weighted-average" core-melt will have consequences of 1.5 x 105 man-rem.
The consequence estimate is 50 man-rem / reactor. On the average, the B&W and CE plants have about 31 caler.dar years of licensed lifetime remaining per plant.
This is roughly 24 years of operational life. Based on the above assumption that one plant will find and correct the problem, the risk reduction estimate is 50 man-rem.
Cos_t Estimate Industry Coct: The costs associated with resolving this item depend on the nature of the solution. A check of the valve operator design is relatively inexpensive. A test to ensure the valves will open will cost significantly more. Finally, if valve operators are found to be insufficiently sized, the cost of replacement will be higher still. In addition, improvements in main-tenance may also be required.
For prioritization purposes, we will assume that a check of design (rather than extensive testing) will be done, and that one plant will be found where the valves would not re-open with a significant differential pressure present.
We will assume further that the motor is strong enough to open the valve and that the problem can be fixed by changing torque limit and bypass switch set-points. Because maintenance error is a relatively minor contributor, we will (for now) not address the issue of improved maintenance.
3.122-6 NUREG-0933 06/30/86 1
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g\ NRC Cost: For each plant affected, 2 staff-weeks should be sufficient to h check the valve design. For the (hypothetical) plant where a problem is found, 6 staff-months should suffice to find a solution. Finally, 6 staff-months plus 2 staff-weeks / plant of NRC time will probably be necessary to impose the re-quirement. 'Thus, for 17 plants, the total cost will be roughly $240,000, assuming that a staff year costs $100,000.
Value/ Impact Assessment Based on a potential risk reduction of 50 man-rem and a cost of $0.24M, the value/ impact score is given by:
3, 50 man-rem
$0.24M
= 208 man- em/$M Other Considerations
- 1. There is no significant ORE associated with the fix for this issue. The valves in question are not exposed to contaminated fluids, since they are in the secondary system.
- 2. There are offsetting savings which could be credited against the expendi-tures above. The cost of a core-melt would be about a billion dollars plus replacement power for the rest of the plant lifetime. In an actu-arial sense, using the accident frequency , estimated above and assuming a 5% annual discount rate, this corresponds to a present worth of about
\
$430,000/ plant. Also, even if core-melt is avoided and if the plant is d ever placed in a situatior,where feed-and-bleed techniques are used, major cleanup will be necessary because of rupture of the quench tank. If cleanup lasts 6 months,- the present worth cost is about $770,000/ plant.
Finally, it should be noted that the Davis-Besse event kept the plant shut down for over three months. The frequency of this situation is about 1.2 x 10 2/RY, which corresponds to an actuarial cost of roughly
$4.6M/ plant.
Obviously, if any of these three considerations were included, the cost-benefit ratio would be favorable indeed. It would be very much in the licensee's financial interest to fix this problem.
- 3. The figures assume-that the feed-and-bleed failure probability is 0.015.
In actual fact, NUREG-1154888 gives the impression that the Davis-Besse operators were rather reluctant to initiate feed-and-bleed. Thus, this figure may be somewhat optimistic. Also, some (CE) plants do not have power-operated relief valves on the primary system and thus cannot use feed-and-bleed techniques.
- 4. Some plants operate with the AFW isolation valves in the closed position.
Thus, these plants will not need an inadvertent isolation to encounter a problem. On the other hand, these plants are more likely to be designed to open under differential pressure or to find the problem by normal p testing.
06/30/86 3.122-7 NUREG-0933
- 5. The discussion has addressed only PWRs; BWRs have analogous systems (HPCI and RCIC) for mitigating loss-of-feedwater events. Moreover, these systems have normally-closed motor-operated isolation valves in the discharge line.
But these valves are tested during normal system testing. In addition, BWRs can rapidly depressurize via the ADS and can use low pressure systems for decay heat removal.
- 6. An OIE Bulletin on the subject of valve operability is being considered.
This may well be sufficient to resolve the issue for most plants. How-ever, some followup action may be appropriate particularly for plants where the viability of feed-and-bleed is doubtful. If such a plant were also susceptible to the common-mode valve problem described here, the core-melt frequency could approach 1 x 10-3
- 7. This issue is related to Item II.E.6.1, "In-Situ Testing of Valves." Al-though II.E.6.1 is also concerned with valve operability, this new item differs in that the potential for commonality is a primary concern. Item II.E.6.1 is geared toward the single failure rate per valve, not the potential for common-mode failures, but is not specific as to which valves or which failure mode.
- 8. This issue is also similar to Issue 87 which concerns the failure of the HPCI steam line isolation valves to close following a break in the line downstream of the valves. These failures are also due to a design problem in which the valve may not have been designed to operate under some over-locked conditions. There may be other systems with valves that are not designed to operate under all likely conditions and therefore a widening of the scope of this issue may be in order.
- 9. It was assumed that the probability of both AFW isolation valves failing to reopen was 32L In some cases (e.g., undersized actuators), this figure may be nearly 100%, which would triple the priority parameters.
However, this would change no conclusions.
CONCLUSION Based on the change in core-melt frequency, this issue should be placed in the HIGH priority category.
ITEM 122.1.B: RECOVERY OF AUXILIARY FEEDWATER DESCRIPTION Historical Background This item addresses Findings 4, 8, and 15 in Sections 5.2.4 and 6.2.4 of NUREG-1154.886 The particular issue deals with a potential inability to remove reactor decay heat due to the second common mode failure discussed above.
This is the excessive delay in recovery of auxiliary feedwater due to diffi-culty in restarting AFW pump steam turbines, if the turbines are tripped.
O 06/30/86 3.122-8 NUREG-0933
/m I
(d Safety Significance Some method of decay heat removal is necessary within 30 minutes after the start of this type of transient in order to prevent core uncovery. The tur-bines tripped about 7 minutes into the event. Thus, 23 minutes were available.
Although it only took 4.5 minutes for a pair of equipment operators to go to the AFW pump rooms and start work, considerable difficulty was experienced in resetting and restarting.the turbines. Thus, it might well have taken longer than 23 minutes to get the AFW pumps in operation. Had other decay heat re-moval techniques (i.e., startup feed pump and primary' side feed and-bleed) also failed, core damage would have resulted.
This issue is applicable to any PWR. However, it is of greatest importance to plants with only steam-driven AFW trains (such as Davis-Besse) and of less importance to plants with one steam-driven train plus one or two motor-driven trains. In addition, non-B&W plants are less susceptible because of their greater water inventory in the steam generators which provides more time before active means of decay heat removal are essential. Davis-Besse is the only re-maining plant with only steam-driven auxiliary feedwater. Thus, this analysis will be geared to the next-most-susceptible plant class: a B&W plant with one steam-driven and one motor-driven AFW train.
Possible Solutions The Davis-Besse event exhibited two problems that led to delay in AFW restart.
pI The first problem was that the turbine overspeed trips had to be manually reset gd requiring plant personnel to be dispatched to the AFW pump rooms. A possible solution is to make the trip resettable from the control room. The trip mecha-nism is usually a latch hook device on the trip-and-throttle valve. A mechani-cal device will unlatch the hook and trip the turbine at a preset speed ('usually 125% of rated). Other signals can be used to trip the latch hook by means of an electrical solenoid. In either case, the hook must be reset manually. The solution, which has been implemented on some BWR RCIC turbines, is to wire the protective circuits into the throttle mechanism rather than the trip solenoid.
The mechanical overspeed trip remains active, but is supplemented by an elec-trical overspeed trip (set at 110%) which can be remotely reset.
The second problem was that the two equipment operators were unsuccessful in their attempts to get the turbines running and were saved by the arrival of an experienced operator. The most obvious solution to this problem would be to require the plant operators to practice going through the procedures of resetting and starting the turbines, assuming a remote reset ,is not provided.
" Hands-on" practice of this task is not now part of operator training.
PRIORITY DETERMINATION Frequency Estimate Problem 1: The affected sequences and cut sets are the same as those for Item 122.1.A except parameter AL, the change in AFW failure probability to be attributed to this item. This is governed by three factors: the proba-bility of a resettable turbine trip, the probability of failure to manually Q reset and restart the turbine, and the probability of failure (in this study) of the one motor-driven AFW train.
06/30/86 3.122-9 NUREG-0933
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l First, we must estimate the probability of a turbine trip either during the auto-start or while running. PRA fault trees model individual components and their failures, but do not normally model the trips of spurious and/or readily resettable trips of concern here. Thus, PRA fault-tree-based estimates are really estimates of the failure rate assuming that the manual reset problem has been fixed. (Also, the turbine-train-only failure rate is remarkably dif-ficult to separate out of most PRA studies.) We will use a value of 3 x 10-2 failure / demand, based on the station blackout calculations for a two-train AFW system in an RRAB memorandum.894 In NUREG/CR-2098,883 112 of the 170 AFW events tabulated were failures of tur-bine rather than motor-driven pumps. Of the 112 turbine events, 40 were trips, usually on overspeed. Thus, given a failure of a turbine-driven AFW pump to operate, there is a 35% chance that a (manual) reset might recover the pump.
Therefore, the failure rate before fixing is (3 x 10-2)(1.35)/ demand or (4.1 x 10-2)/ demand.
We must now estimate the change in turbine failure rate due to elimination of the need for manual reset. In the Davis-Besse event, the operators were able to reset the two turbines in 4.32 and 4.77 minutes (but not get them running),
which was about one-fifth of the 23.4 minutes available before core uncovery. ass One would expect that, for a straightforward task such as resetting and re-starting a turbine, the time needed would be described by a reasonably symmet-rical distribution centered about an average time. Here, the 4.5-minute aver-age time of the two unsuccessful resets at Davis-Besse is probably a reasonable estimate of a general mean time for an experienced operator to successfully complete the task. This number is also consistent with oral communications we have had with operations staff at two other plants and with a walk-through of the procedure at Davis-Besse by NRC staff. However, we have no direct informa-tion about the width of the distribution--the minimum and maximum time needed for completion. Thus, we will use a pragmatic approach. We will keep the peak of the distribution at 4.5 minutes and fix it at zero at time equals zero.
Further, we will use the single-event Poisson distribution which will extend out to infinity in the positive direction. The formula is P(t) = At exp(-At)
The peak of the distribution is at t = 1/A so we will use A = 1/4.5 minutes = 0.22. The probability of not resetting the turbine before 23.4 l
minutes is obtained by integrating this formula from 23.4 minutes to I infinity. The integral is:
P(t > t ) = (1 + Atg ) exp (-At g) l g
= 0.036
= 3.6%
Again, this approach is pragmatic rather than rigorous--the formula is appro-priate for randomly distributed events, which this really is not. In the actual event at Davis-Besse, it is evident that the operating crew worked as fast as they could. It is also evident that the task of resetting and restarting the turbines was far from smooth; many things went wrong. Moreover, things might well not be easy and straightforward in another similar event. Nevertheless, a I factor of five margin in the time actually taken is significant. Thus, 3.6%
does not seem unreasonable in spite of the rather sparse mathematical basis.
06/30/86 3.122-10 NUREG-0933 1
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[ \ In addition, there is a finite probability that plant operators will encounter
( ,) difficulty in moving through the plant and entering the AFW pump rooms due to locked doors, etc. To account for this, we will add a 1% probability of an in-surmountable difficulty in reaching the turbines (based on the calculations in Issue 122.3) to get a total probability of failure to reset of 0.046.
AL can now be estimated. First, the change in the turbine-driven train's failure rate is:
g 4.1 x 10 2 failures 3g 0.35 turbine trips _3 g 0.046 failure to reset demand Total failures Turbine trip 3
= 6.6 x 10 4/ demand In addition, we must estimate the unavailability of the motor-driven train.
The RRAB memorandum 894 gives a " typical" AFW system unavailability of 10 3/
-demand for a two-train system. Such a figure includes common-mode failures and common component failures in addition to the individual train failures. For our purposes, we will assume that the common-mode and common-component contri-butions are small and thus the turbine train contribution enters as a multipli-cative factor. The non-turbine failure rate is then 10-3/(3 x 10 1) or 0.033.
Giving credit for the motor-driven train, if AC power is available, AL = (6.6 x 10 4)(0.033) = 2.2 x 10 5 If AC power is not available, v
AL' = 6.6 x 10 4 One more figure is needed. Since the turbine-driven AFW pump is especially significant for loss of all AC ower (station blackout), a diesel unavail-ability is needed. NUREG-10328 0 gives a range of 1.1 x 10 3 to 6.8 x 10 3 for a one-out-of-two diesel configuration. We will use 2.7 x 10 3, the middle of this range.
Cut sets can now be calculated:
T iM*LOPNRE*aL*U 4.1 x 10 9/RY T iM*LOPNRE* DIESELS *AL' 2.0 x 10 8/RY T2M*aL*U 2.1 x 10 7/RY Ta*M*PCSNR*aL*U 2.1 x 10 9/RY Total AF = 2.4 x 10 7 core-melt /RY For 9 PWRs with two-train AFW systems, this frequency is 2.2 x 10 6 core-melt / year.
Problem 2: In the first problem, it was assumed that the only question was the O time available for a qualified operator to locally reset a tripped AFW turbine.
The fact that neither of two equipment operators was able to get the turbines 06/30/86 3.122-11 NUREG-0933
running at Davis-Besse strongly suggests that the probability of failure is nearly unity over the course of a half-hour, if the individuals involved have never performed this task before. (This task is generally not part of an operator's training.) In general, during off-shifts, experienced personnel are present in very limited numbers. In a future event, the more experienced per-sonnel are likely to be busy with other tasks (e.g., getting diesels started),
and a less experienced operator may once again be faced with the task of resetting and restarting AFW turbines.
This second problem is not amenable to the exponential time calculations of Problem 1, since the average time needed for inexperienced personnel is likely to be far in excess of 30 minutes. Thus, we will arbitrarily assume that, should an event occur during the evening, night, or weekend shifts (76% of the time), there is a 50% probability that an AFW turbine trip reset will be assigned to an inexperienced operator who is at most 10% likely to succeed in getting the turbine running in the required time. Thus, the change in the prob-ability of failure to restart the turbine becomes (0.76)(0.50)(0.90) = 0.342.
For this problem, the change in the turbine-driven train's unavailability is:
4.1 x 10-2 failure 3E 0.35 turbine trip 3 E 0.046 failure to restart E 3 demand failure turbine trip
= 4.9 x 10 3/ demand Giving credit for the motor-driven train as before:
AL = (4.9 x 10 1) (0.033) = 1.6 x 10 4 (AC power available)
AL' = 4.9 x 10 3 (AC power not available)
Cut sets can now be calculated:
T 1M*LOPNRE*AL*U 3.1 x 10 8 T i M*LOPNRE*0IESELS*aL' 1.7 x 10 7 T2M*3L*U 1.6 x 10 8 Ta*M*PCSNR*aL*U 1.6 x 10 8 Total AF = 1.8 x 10 6 core-melt /RY For 9 PWRs with two-train AFW systems, this frequency is 1.6 x 10 5 core melt / year.
Consequence Estimate The consequence estimate is the same as that for Item 122.1.A. The " weighted-average" core-melt will have consequences of 1.5 x 10 5man-rem. The 9 PWRs with two-train AFW systems have about 250 calendar years of collective license lifetime remaining. This is roughly 189 years of operational life.
O 06/30/86 3.122-12 NUREG-0933
/"g . Problem 1: The consequence estimate is (2.4 x 10 7)(1.5 x 105)(189) man-rem =
Q 7 man-rem. l Problem 2: The consequence estimate is (1.8 x 10 8)(1.5 x 105)(189) man-rem =
51 man-rem.
Cost Estimate Problem 1: Changing the turbine trip logic on a safety-related system is likely to require 6 staff-months of effort per plant, even if no major pro-curement is needed. In addition, at least 3 staff-months of generic work plus a week of effort on each plant will be required of the NRC staff. The total cost for the 9 PWRs with 2 AFW trains (excluding Davis-Besse) is thus at least
$0.5M.
Problem 2: Having operators practice the task of resetting and manually star-ting AFW turbines is relatively inexpensive. (If, after the first time, more than half an hour of the operator's time is needed, there is little point in the exercise.) However, this is a continuing expense. We will assume one staff-month / plant of administrative effort to set the program up plus two staff-weeks / year thereafter of actual practice. Assuming a 5% discount rate and an average remaining life of 28 calendar years, this is about $620,000 total for 9 plants. NRC costs are again likely to be one staff-month of ge-neric work plus 1 staff-week / plant, or about $26,000. The total cost is roughly
$650,000, h
Value/ Impact Assessment Problem 1 The value/ impact score is given by:
3; 7 man-rem
$0.5M
= 14 man-rem /$M Problem 2 The value/ impact score is given by:
3 ; 51 man-rem
$0.65M
= 78 man-rem /$M Other Considerations
- 1. There is no significant ORE associated with the fix for this issue. The valves in question are not exposed to contaminated fluids, since they are in the secondary system.
- 2. There are offsetting savings which could be credited against the expendi-N tures above. The cost of a core-melt would be about $1 billion plus 06/30/86 3.122-13 NUREG-0933
replacement power for the rest of the plant lifetime. In an actuarial sense, using the accident frequencies estimated above and assuming a 5%
annual discount rate, this corresponds to a present worth or $6,000/ plant.
Also, even if a core-melt is avoided and the plant is ever placed in a situation where feed-and-bleed techniques are used, major cleanup will be necessary because of rupture of the quench tank. If cleanup lasts six months, the actuarial cost has a present worth of $10,000/ plant.
- 3. The figures assume that the feed-and-bleed failure probability is 0.015.
In actual fact, NUREG-1154888 gives the impression that the Davis-Besse operators were rather reluctant to initiate feed-and-bleed.
Thus, this figure may be somewhat optimistic. Also, some (CE) plants do not have power-operated relief valves on the primary system and thus can-not use feed-and-bleed techniques. Raising the feed-and-bleed failure probability to 0.1 would put this issue into the high priority range.
- 4. Some plants may have still other means of decay heat removal (e.g. the high head service water system at Oconee). For these plants, the figures would have to be adjusted downward.
- 5. These figures should not be used for BWR HPCI and RCIC systems. The BWR systems generally have a greater number of trips and an elaborate isola-tion system.
- 6. The calculations above are based on an AFW system with one motor-driven and one turbine-driven train. A plant such as Davis-Besse with only two turbine-driven trains will be significantly more susceptible to this issue because whatever tripped the first turbine may well trip the second also.
Other plants which originally were equipped with only turbine-driven trains include Turkey Point 3 and 4 and Haddam Neck. The Turkey Point units share three turbine-driven AFW trains and also have each installed a motor-driven train. Haddam Neck has two turbine-driven trains and has installed one (manual start) motor-driven train. The availability and surveillance requirements for the new motor-driven trains on these plants have not been added to the plants' technical specifications and they are as yet not capable of being powered from onsite emergency power. Never-theless, given the presence of these diversely powered trains, these plants are not likely to need special treatment for this issue.
CONCLUSION This issue is of high priority for those plants which cannot remove decay heat by feed-and-bleed or other alternative means. Thus, it should be subsummed into Issue 122.2 for such plants. For the remainder, based on the figures above, this issue should be placed in the MEDIUM priority category.
O 06/30/86 3.122-14 NUREG-0933
ITEM 122.1.C: INTERRUPTION OF AUXILIARY FEEDWATER FLOV
[mV) DESCRIPTION Historical Background This item addresses Finding 6 in Section 5.2.2 of NUREG-1154.sse The particu-lar issue deals with a potential inability to remove reactor decay heat because
~
of the interruption of all auxiliary feedwater flow due to the first common mode failure discussed above. This is the closing of the AFW pump discharge isolation valves. This is related to Issue 122.1.A, which deals with another problem that prevented the isolation valves from reopening.
Safety Significance The definitions 8s of this issue is ambiguous in that the full title, " Inter-ruption of Auxiliary Feedwater Flow due to Failures in Steam and Feed Line Break Accident Mitigation Features (e.g., SFRCS)," refers to the second failure described under 122.1, but the bases presented are Section 5.2.2 and Finding 6 of NUREG-11548ss which refer to the first failure (i.e., of main, not auxil-iary, feedwater). We will address both in this analysis.
The first sub-issue is the spurious closure of the MSIVs, in this case as a result of a turbine trip. Most plants of recent design are equipped with turbine-driven main feedwater pumps. Closure of the MSIVs will shut off all f~3 feedwater flow. Moreover, once MSIVs are closed, the reopening of these I valves is a rather elaborate procedure. The loss of main feedwater is not k.)t easily recoverable.
The second subissue is the isolation of auxiliary feedwater. This is done in the event of a steam line break within containment to prevent exceeding the containment design pressure. The containment is designed to accommodate the initial blowdown of a steam generator. If feedwater to the affected steam gen-erator is not shut off, the boil-off due to decay heat will continue to dump steam to the containment. However, in a transient involving loss of main feed-water but no steam line break, shutting off AFW flow is very undesirable. It must also be remembered that loss-of-feedwater events are far more frequent than steam line breaks.
Possible Solutions Inadvertent MSIV closure has in the past been considered a relatively rare transient. In the particular case of the Davis-Besse transient, the steam gen-erator level sensors had been replaced by a new type of transmitter.686 The rapid closure of the turbine stop valves sent a pressure wave up the steam lines back to the steam generators. This phenomenon is not new; it is rou-tinely allowed for in the analysis of BWR transients where the ' reactor core is directly sensitive to the pressure pulse. However, the new trar.smitters were of a design that did not dampen out the pressure pulse, which cased them to trip. A possible solution would be to add some damping to the level signal at those plants where this has proven to be a problem.
/~N 06/30/86 3.122-15 NUREG-0933
The inadvertent isolation of AFW flow appears to be primarily a human factors prcblem associated with the controls layout. This could be solved by a rede-sign of this portion of the control panel. If on further study it appears that spurious isolations are occurring because of hardware problems, other actions (e.g., possibly using high containment pressure in a logical "and" with low steam generator pressure) might be necessary. In addition, the ;uestion of whether an operator should anticipate automatic actuations or simply observe and confirm them should be addressed in the long term.
This item appears to be associated with B&W plants. The isolation logic and AFW control is quite different for the other PWR vendors. (CE-designed plants may be susceptible to the first subissue.)_
PRIORITY DETERMINATION Frequency Estimate The affected sequences and cut sets are the same as those for Item 122.1.A with the exception of the parameter L which is redefined as follows:
L - This is the failure rate of the auxiliary feedwater system.
The RRAB memorandum 894 gives 10-3/ demand as " typical" for a two-train system (offsite power available) and 1.8 x 10 5/ demand as " typical" for a three-train system.
The first subissue, inadvertent MSIV closure, has the effect of turning the Ta- hitiated transients into T2 -initiated transients. (T1 transients are unatfected). If every transient led to MSIV closure (as NUREG-1154,886 Sec-tion 5.11 seems to imply), the parameters and sequences are straightforward:
AT 2 = (5.7 - 0.64) = 5.06 ATa = -S 7 for plants with a two-train AFW system:
AT2M*L*U 7.6 x 10 5 AT3*M*PCSNR*L*U -9.5 x 10 8 Net change, AF = 7.6 x 10 5/RY For plants with a three-train AFW system:
AT2M*L*U 1.4 x 10 6 AT3*M*PCSNR*L*U -1.7 x 10 9 Net change, AF = 1.4 x 10 6/RY The second subissue, AFW isolation, affects parameter L. The change in L is composed of two factors: the change in the probability of spurious isolation and the probability of failure to reopen on demand. As discussed in 06/30/86 3.122-16 NUREG-0933
/
V) issue 122.1.A, we will assume a 5% minimum likelihood of spurious AFW isolation and assume further that another plant with a high (e.g. , 20%) likelihood' exists.
The second factor is the failure of the isolation valves to reopen on demand.
We will assume that Item 122.1. A has been addressed independently and that this failure probability is now governed by the failure of an operator to diag-nose and correct the problem. The operator failure rate for such a situation is not' independent of the spurious actuation error described above. We will assume, based on judgment, that 95% of the time the operator will correct the
. error by resetting the inadvertent isolation and reopening the isolation valves.
For the more realistic (5% inadvertent isolation probability situation, the cut sets become:
T tM*LOPNRE*AL*U 4.7 x 10 7 T2M*Al'U 2.4 x 10 5
'Ta*M*PCSNR*AL*U 2.4 x 10 7 Total AF = 2.5 x 10 5/RY For.the more extreme (20%) case, this change in care-melt frequency would be four times this, or 9.9 x 10 5 Consequence Estimate In\
V The consequence estimate is the same as that for Item 122.1.A. The " weighted-average" core melt will have consequences of 1.5 x 105 man-rem.
Cost Estimate The core-melt frequencies are'in a range where costs that are within reason will not affect priority assignments. Consequently, no cost analysis has been made.
Value/ Impact Assessment Without a detailed design examination, it is not possible to determine exactly i how many plants are affected. The B&W plants have an average of 29.5 calendar-years (22 operational years) of lifetime left. Priority parameters are:
1 Subissue 1 Subissue 2 Man-rem / reactor 250 80 Core-melt /RY 7.6 x 10 5 2.5 x 10 5 Other Considerations
- 1. There is no significant ORE associated with the fix for this issue. The p valves in question are not exposed to contaminated fluids since they are in the secondary system.
06/30/86 3.122-17 NUREG-0933
l l
l l
i
- 2. The figures assume that the feed-and-bleed failure probability is 0.015. In actual fact, NUREG-1154 88s gives the impression that the Davis-Besse operators were rather reluctant to initiate feed-and-bleed.
Thus, this figure may be somewhat optimistic, which would raise the priority scores still higher.
- 3. The two subissues were evaluated separately above because they involved two separate failures in the Davis-Besse event. Nevertheless, it should be noted that both involved the SFRCS. In essence, one control system apparently has the capability to shut off both main feedwater (by MSIV closure) and auxiliary feedwater. Although two distinct failures were involved at Davis-Besse, there may well be a single failure within the SFRCS which could do both. Deterministic evaluations of this system should recognize the seriousness of such a failure mode.
CONCLUSION Based on the core-melt frequency figures above, this issue should be placed in the HIGH priority category.
ITEM 122.2: INITIATING FEED-AND-BLEED DESCRIPTION Historical Background This issue deals with the adequacy of emergency procedures, operator training, and available plant monitoring systems for determining the need to initiate feed-and-bleed cooling following loss of the steam generator heat sink. It is based upon Findings 10, 17 and 18 in Sections 6,1.1 and 6.1.2 of NUREG-1154.886 Essentially, the operators were reluctant to take the rather drastic step of initiating feed-and-bleed cooling, probably because they believed restoration of the AFW system was imminent. The fact that feed-and-bleed cooling releases primary coclant to the containment (implying an extensive shutdown for the pur-pose of decontamination) may also have influenced their actions. Finally, the normal control room instrumentation was inadequate to clearly inform the opera-tors that feed-and-bleed was called for. The SPDS which would have displayed the necessary information was not operable.
The reactor vendors have provided their customers with feed-and-bleed proce-dures. Feed-and-bleed capability is not currently specifically required by the NRC although the techniques, benefits, and costs are being evaluated as part of USI A-45. Basically, feed-and-bleed cooling is a method of last resort which can avert core damage if main and auxiliary feedwater is lost and other methods of deccy heat removal are unavailable. For plants licensed without a PORV, the lack of feed-and-bleed capability was a significant issue and the need for a highly reliable AFW system was emphasized.
Safety Significance PRAs give considerable credit for feed-and-bleed cooling. A failure rate of one or two percent is a typical assumption. However, the Davis-Besse event 06/30/86 3.122-18 NUREG-0933
chronology leaves an impression that this failure probability may be overly g optimistic.
In addition, it should be~noted that, depending on specific plant design, there may be a fairly short time period in which feed-and-bleed cooling will be successful. If the plant operators delay too long before initiating feed-and-bleed cooling, their error may not be retrievable by later action.
This issue applies to all plants which can use feed-and-bleed techniques.
This is all PWRs except for a few CE-designed plants which have no pressurizer PORVs.
i Possible Solutions The solution is a matter of emphasis on safety vs. operation, training in existing procedures, and possibly an upgrading of instrumentation at certain sites. In addition, the procedures themselves could be upgraded to make the criteria for initiation of feed-and-bleed cooling more direct and unambiguous, leaving less room for operator reluctance. (For example, in the case of j
Davis-Besse, basing the initiation of feed-and-bleed on hot leg temperature rather than on steam generator parameters has been suggested.) Here, we will concentrate on ensuring that existing procedures are.followed. The general technical aspects of feed-and-bleed decay heat removal will be addressed under i
USI A-45.
- PRIORITY DETERMINATION Frequency Estimate The question of interest is, what is the change in core-melt frequency if the failure probability of feed and-bleed cooling (U) is changed? NUREG/CR-165954 and NSAC-60889 assume a failure probability of 0.015 for.non-ATWS sequences (RSSMAP parameter "HPMAN") and 0.10 for the (higher stress) ATWS sequences
("HPMAN1"). The operators' performance during the Davis-Besse event leaves a strong impression that.these figures are too low. We will assume, based purely on judgment,.that failure probabilities of 0.10 for non-ATWS sequences and 0.50 for ATWS sequences are more reasonable estimates.
In making the calculations, the parameters were the same as in Issue 122.1.A, except:
! a. The frequency of loss of main feedwater transients T ,(momentary 2 and sustained) was set at 2.13/ year, based on NSAC-60.889 b.
The'AFWfailurep94robability (L) was set as follows, based on the RRAB memorandum:
3 Offsite Power Available No Offsite Power 3-train AFW 1.8 x 10 5 5.1 x 10 5 2-train AFW 1.0 x 10 3 1.7 x 10-3 0 In addition, the computerized RSSMAP54 analysis was changed as fnllows:
06/30/86 3.122-19 NUREG-0933
, - . . , - . . . , . , . . . , , , , - , - - - - - . - . , _ , - - . , . . . . , , , , . , , ~ . . - , . , , , - , . . _ , , , , . . - , - , . - - - - , . . - . , , . .
- a. The probability of loss of onsite power (8 )3was changed to 1.3 x 10-3, a figure more representative of a twin diesel system.
(0conee uses hydroelectric generators for emergency power.)
- b. Oconee's capability of feeding the steam generators with the High Head Service Water System was disabled (HHMAN = 1.0).
A series of computer calculations was performed, in an attempt to obtain both the "best" answ?r and some information as to the sensitivity of the answer to a variety of conditions.
Calculation AF (Core-melt /RY) 3-train AFW system HPMAN raised to 0.1 HPMAN1 raised to 0.5 3.3 x 10 5 3-train AFW system, HPMAN raised to 0.1 ATWS sequences unchanged 9.2 x 10 6 2-train AFW system HPMAN raised to 0.1 HPMAN1 raised to 0.5 1.0 x 10 4 2-train AFW system HPMAN raised to 0.1 ATWS sequences unchanged 8.1 x 10 5 Test case, original RS$ MAP parameters.
HPMAN raised to 0.1 ATWS sequences unchanged 2.0 x 10 5 Clearly, the change in the feed-and-bleed failure probability has a strong effect on core-melt frequency. The figures span the decade from 10 5 to 10 4 We will use the first calculation (3.3 x 10 5) bearing in mind that the figure for a plant with a two-train AFW system will probably be greater. In addi-tion, it should be noted that even a partial solution will make a significant reduction in core-melt frequency.
There are 55 operating PWRs, with an aggregate of about 1700 calendar years or 1300 operationa' years of lifetime remaining. Thus, the frequency estimate is (3.3 x 10 5)(,n) core-melt / year or 1.8 x 10 3 core-melt / year.
Consequence Estimate The consequence estimate is the same as that for Item 122.1.A. The " weighted-average" core-melt will have consequences of 1.5 x 105 man-rem. For 55 plants with a combined remaining operation life of 1300 years, the consequence esti-mate is approximately 6,500 man-rem.
O 06/30/86 3.122-20 NUREG-0933
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Cost Estimate LJ The fix for this issue is likely to be procedural in nature, with upgrades in equipment more likely to be done under USI A-45. We will assume that 6 staff-months / plant will suffice for refresher training on these procedures. NRR costs are likely to be on the order of 6 staff-months of generic effort plus 2 staff-weeks per licensee. For 55 operating PWRs, this is roughly $3M.
Value/ Impact Assessment Based on a risk reduction of 6,500 man-rem and a cost of $3M, the value/ impact score is given by:
3 = 6,500 man-rem 53M
= 2,167 man-rem /$M Other Considerations
- 1. For a plant with a two-train AFW system, the per-reactor and per-RY figures will be roughly three times as large.
- 2. This issue does not involve ORE.
TN 3. There is an offsetting saving which could be credited against the expendi-f 4 tures above. The cost of a core-melt would be about a billion dollars U plus replacement power for the rest of the plant lifetime. In an actuarial sense, using the accident frequencies estimated above, assuming a 5% annual discount rate and subtracting off the feed-and-bleed cleanup costs wh'ich would reduce the core-melt costs, this corresponds to about a present worth of $1.2M/ plant.
- 4. In contrast to the saving associated with averting a core-melt, an unneces-sary use of feed-and-bleed will result in major cleanup costs. If half the uses of feed-and-bleed are unnecessary and a cleanup lasts six months, the actuarial cost shows a present worth of roughly $400,000/ plant (based on a residual frequency of unnecessary use of feed-and-bleed of 5 x '
10 4/RY).
CONCLUSION Based on the' figures above, this issue should be placed in the" HIGH priority category.
ITEM 122.3: PHYSICAL SECURITY SYSTEM CONSTRAINTS DESCRIPTION Historical Background Q This particular issue arose out of Finding 9 in Section 3.6 of NUREG-1154,8ss which states:
06/30/86 3.122-21 NUREG-0933 r n e.- ,. - -,.---, ,,, -,,-
"The locked doors and valves in the plant had the potential for significantly hampering operator actions taken to compensate for equipment malfunctions during the event and were a significant concern to the equipment operators."
In the Davis-Besse event, the operators were able to reach the AFW pump room with no reported difficulty. There were difficulties in resetting and restarting the turbines and in opening the isolation valves, but these were not related to locking devices.
Safety Significance Barriers and locks are present for purposes of physical security, as the title of this issue implies. In addition, barriers are provided for other purposes, such as personnel protection, fire zone isolation and flood protection. Valves are locked not only for security reasons, but also because inadvertent opening of these valves may have economic or safety consequences. The presence of the locking devices and barriers must strike a balance between these purposes and the fact that these devices may impede free movement in the plant and some local operations during an emergency. It should be noted that the control boards in the control room are also liberally supplied with keylock switches. This issue applies to all reactors.
Possible solution The possible solution for this issue is to completely evaluate the net effect of a given barrier on plant safety and either remove it or (in extreme cases) provide an alternate means of entrance (with its own locks), should the analy-sis so indicate.
PRIORITY DETERMINATION This issue is not new. Issue 81 evaluated the impact of locked doors and bar-riers on safety, considering the frequency of a need for entry into the plant, the likelihood of procedural error (e.g., wrong key), and probability of suc-cessful forcible entry in a timely fashion. The conclusion was a " drop" priority assignment.
Issue 81 considered only non-security barriers. A barrier that was installed for security reasons is not as likely to be forcibly penetrated in a few min-utes. Moreover, the scenario here is slightly different than that of Issue 81.
It should be noted, however, that the Davis-Besse experience confirms some uf the assumptions of the Issue 81 evaluation since there were in fact no problems with locked doors or valves.
Frequency Estimate We will estimate frequency based on a loss of main feedwater event consistent with Issue 122.1.A. The frequencies and probabilities are: non-recoverable loss of main feedwater (0.67/RY), failure of auxiliary feedwater (use 10 3 for a " typical" two-train system and 1.8 x 10 5 for a " typical" three-train system), and failure of feed-and-bleed cooling (0.015).
We will further assume that a locked barrier may prevent entry into the auxil-f ary feedwater pump room (s) and that such entry could recover the AFW system.
06/30/86 3.122-22 NUREG-0933
N g This is a high stress situation. Thus, we will assume that there is a 10%
% chance of human error (e.g., wrong key) and a 10% chance of non-recovery. (The chance of mechanical lock failure estimated in Issue 81 is 0.001.) We will not assume credit for forcible penetration.
We will not consider the padlocks and chains on the valve wheels, in view of the existence of bolt cutters and the fact that there will be two or three redundant trains. The result is a change in core-melt frequency of 10 7 for plants with 2 AFW trains and 1.8 x 10 8 for plants with 3 AFW trains.
Consequence Estimate The consequence estimate is the same as that for Item 122.1.A. The " weighted-average" core-melt will have consequences of 1.5 x 105 man-rem. Assuming 30 years of remaining operational life for plants with 2 AFW trains, the conse-quence estimate is (10 7)(1.5 x 105)(30) man-rem / reactor or approximately 0.45 man-rem / reactor. For plants with 3 AFW trains, the consequence estimate is (1.8 x 10 8)(1.5 x 105)(30) or approximately 0.01 man-rem / reactor.
Cost Estimate Issue 81 estimated a one-time evaluation of existing locked doors to cost
$200,000. We wi_11 use this as a minimum per plant cost, recognizing that an adverse finding will incur labor and equipment costs that may be much larger.
r Value/ Impact Assessment (v ]) 2 AFW Trains Based on a risk reduction of 0.45 man-rem / reactor, the value/ impact score is given by:
3 _ 0.45 man-rem / reactor
$0.2M/ reactor
= 2.25 man-rem /$M 3 AFW Trains '
Based on a risk reduction of 0.01 man-rem / reactor, the value/ impact score is given by:
3 _ 0.01 man-rem / reactor
$0.2M/ reactor
= 0.05 man-rem /$M Other Considerations The analysis is based on the PWR design. It is not expected that a BWR design would be greatly different from that of a three AFW-train PWR, given the ability
- of HPCI, RCIC, and the ADS low pressure ECCS to mitigate transients.
CONCLUSION Based upon the figures above, this issue should be placed in the LOW priority category.
06/30/86 3.122-23 NUREG-0933
REFERENCES
- 54. NUREG/CR-1659, " Reactor Safety Study Methodology Applications Program,"
U.S. Nuclear Regulatory Commission, 1981~.
- 64. NUREG/CR-2800, " Guidelines for Nuclear Power Plant Safety Issue Prioritization Information Development," U.S. Nuclear Regulatory Commission.
885. Memorandum for H. Thompson from D. Crutchfield, " Potential Immediate Generic Actions as a Result of the Davis-Besse Event of June 9,1985,"
August 5, 1985.
886. NUREG-1154, " Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9, 1985," U.S. Nuclear Regulatory Commission, July 1985.
887. Memorandum for T. Speis from H. Thompson, "Short Term Generic Actions as a Result of the Davis-Besse Event of June 9, 1985," August 19, 1985.
888. Mhmorandum for H. Denton from T. Speis, " Adequacy of the Auxiliary Feedwater System at Davis-Besse," July 23, 1985.
889. NSAC-60, "A Probabilistic Risk Assessment of Oconee Unit 3," Electric Power Research Institute, June 1984.
890. NUREG-1032, " Evaluation of Station Blackout Accidents at Nuclear Power Plants," U.S. Nuclear Regulatory Commission, (Draft) May 1985.
l 891. Letter to T. Novak (NRC) from R. Crouse (Toledo Edison Company),
l December 31, 1981.
892. NUREG/CR-2770, " Common Cause Fault Rates for Valves," U.S. Nuclear Regulatory Commission, February 1983.
893. NUREG/CR-2098, " Common Cause Fault Rates for Pumps," U.S. Nuclear Regu-latory Commission, February 1983.
894. Memorandum for 0. Parr from A. Thadani, " Auxiliary Feedwater System -
CRGR Package," November 9, 1984.
O 06/30/86 3.122-24 NUREG-0933
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[U ISSUE 124: ' AUXILIARY FEEDNATER SYSTEM RELIABILITY DESCRIPTION Operating experience as well as staff and industry studies indicate that AFW systems continue to fail at a high rate. These studies also indicate that plants with similar AFW system reliabilities (as calculated in accordance with the SRP11 guidance) do not necessarily exhibit similar AFW system availabilities. Based on these studies and on engineering judgment, the staff concluded that the PWR AFW system reliabilities calculated in accordance with the SRPit guidance'may represent the relative reliability of AFW system hardware configurations for various plants, but do not represent the real availability of these crucial safety systems.914 In order to ascertain a high level of AFW system reliability and availability, the staff proposed a requirement that all operating plants demonstrate by PRA
, that their AFW systems are at least as reliable as 10 4 unavailability / demand 1
after accounting for: (a) AFW system support systems, (b) common cause failures, or (c) operator errors. As input to the PRAs, each utility should use its plant-specific data if available. Such plant-specific data will reflect design faults, poor maintenance practices, and inadequate testing and surveillance and will indicate how well a particular plant is being operated, thereby identifying (v those plants that will need improvements.
CONCLUSION Because of the significance of the AFW system in reducing core-melt frequency,
- he staff has determined that all PWRs should meet the reliability criterion saecified in SRP11 Section 10.4.9. This SRP Section was not applied to operat-inq reactors. In order to achieve and maintain a high degree of reliability for the AFW system or alternate decay heat removal, the following should be completed: (1) PWR licenses and applicants should demonstrate, using reliabil-
! ity analyses, that the AFW system is of high reliability (10 4 to 10 5 unavail-ability per demand); (2) the staff is to review the reliability analyses and/or any necessary system modifications and procedural or maintenance changes. As a result of (1) and (2) above, the staff will determine whether it is necessary to require that plants upgrade their AFW systems to the s.afety-related standards.
Based on the staff evaluation of W, CE, and B&W plants' AFW systems (NUREG-0611,93 NUREG-0635,95 B&W plant SERs), the staff has determined that the AFW system of the following plants are not sufficiently reliable and should be upgraded:
t Prairie Island 1, Prairie Island 2, ANO-1, ANO-2, Fort Calhoun, Crystal River, and Rancho Seco. Thus, the solution to this issue has been identified.
REFERENCES
- 11. NUREG-0800, " Standard Review Plan," U.S. Nuclear Regulatory Commission.
] 93. NUREG-0611, " Generic Evaluation of Feedwater Transients and Small-Break
/ Loss-of-Coolant Accidents in Westinghouse Designed Operating Plants,"
U.S. Nuclear Regulatory Commission, January 1980.
06/30/86 3.124-1 NUREG-0933
- 95. NUREG-0635, " Generic Assessment of Small Break Loss-of-Coolant Accidents in Combustion Engineering Designed Operating Plants," U.S. Nuclear Regulatory Commission, January 1980.
914. Memorandum for H. Thompson and T. Speis from R. Bernero, " Request for Comments on Draft CRGR Package with Requirements for Upgrading Auxiliary Feedwater Systems in Certain Operating Plants," October 3, 1985.
O O
06/30/86 3.124-2 NUREG-0933
p i 1 V
ISSUE 125: DAVIS-BESSE LOSS OF ALL FEEDWATER EVENT OF JUNE 9, 1985 - LONG TERM ACTIONS On June 9, 1985, Davis-Besse'had a partial loss of feedwater while operating at 90% power. Following a reactor trip, the loss of all feedwater occurred. The two OTSGs became dry and were ineffective as a heat sink. Consequently, the RCS pressure increased indicating a lack of heat transfer from the primary to secondary coolant systems. The PORV automatically opened and closed twice during the event upon reaching the approximate pressure setpoints; it' opened a third time,.but did not re-close for some unknown amount of time. The delayed response to close the third time aggravated the recovery of the event and allowed a rapid depressurization of the RCS. A staff report on the event was published in NUREG-1154888 and an EDO memorandum 895 identifying NRR' action items was issued on August 5, 1985. For prioritization purposes, the items were broken down into two groups: (I) Issues raised in NUREG-1154 and the EDO memorandum and (II) Other Issues. These items are prioritized separately below.
ITEM 125.I.2: PORV RELIABILITY The PORV common to most PWRs (with the exception of CE 3410 and 3800 Mwt plants and ANO-2) is designed to limit system pressure if a transient recovery exceeds the capability of the pressurizer spray system. Davis-Besse has a solenoid-p
- )i U
controlled PORV. However, many other PWRs have PORVs that are operated pneu-matically (instrument air or nitrogen). Both designs have the same purpose.
The PORV is designed to receive an actuation signal to open from the pressurizer pressure instrumentation at a design setpoint (typically 2425 psig) in order to prevent reactor pressure from rising and activating the code safety valves.
If a PORV is used for feed-and-bleed, it can either be: (1) set to stay open by the operator dropping the setpoint low enough sucn that the valve will remain open until reaching the lower setpoint for LPIS or RHR initiation, or (2) cycled ,
open and closed many times, should there be a need for feed-and-bleed. Option 1 appears to be the more common practice.
PORVs are also used in other functions such as mitigating SGTR accidents, LTOP, or RCS venting. Its performance is required for plant protection and accident mitigation.
The following is the evaluation of the four parts of this issue.
ITEM 125.I.2.A: NEED FOR A TEST PROGRAM TO ESTABLISH RELIABILITY OF THE PORV DESCRIPTION Historical Background This issue was identified as Item 9c in the EDO memorandum 895 and is based on O' Finding 13 and Section 5.2.8 of NUREG-1154.8ss 06/30/86 3.125-1 NUREG-0933
Safety Significance Although the PORV can be used successfully in recovering from certain plant transients, there has been no suitable test program established to verify its reliability.ss6 This issue affects all PWRs that can use PORVs.
CONCLUSION The need for improving the reliability of PORVs and block valves, in light of plant protection and accident mitigation requirements, is being addressed in the resolution of Issue 70, "PORV and Block Valve Reliability." Revised licens-ing criteria may be developed, if needed, to include testing requirements.896 Therefore, this issue is covered in Issue 70.
ITEM 125.I.2.B: NEED FOR PORV SURVEILLANCE TESTS TO CONFIRM OPERATIONAL READINESS DESCRIPTION Historial Background This issue was identified as Item 9d in the EDO memorandum895 and is based on Finding 13 and Section 5.2.8 of NUREG-1154.886 Safety Significance The review of the PORV maintenance and operating history reveals that the me-chanical operation of the valve had not been tested and that the valve had not otherwise been operated for over 2 years and 9 months prior to the June 9, 1985 event. Therefore, it seems that there exists a need for surveillance tests to confirm operational readiness. This issue affects all PWRs that can use PORVs.
C_0NCLUSION The number of times that PORV/ Block Valves are used during a typical fuel cycle will be reviewed in the resolution of Issue 70, "PORV and Block Valve Reliabil-ity," in order to determine if a surveillance program should be initiated to confirm operational readiness.896 Therefore, this issue is covered in Issue 70.
ITEM 125.I.2.C: NEED FOR ADDITIONAL PROTECTION AGAINST PORV FAILURE DESCRIPTION Historical Background This issue was identified as Item 9e in the ED0 memorandum89s and is based on Sections 5.2.8 and 6.2.1 of NUREG-1154.886 Safety Significance The PORV will receive an actuation signal from pressurizer pressure instrumenta-tion at a design setpoint (typically 2425 psig) to open in order to prevent 06/30/86 3.125-2 NUREG-0933
/m reactor pressure from activating the code safety. valves. After the opened PORV V) has reduced the pressure sufficiently to reach its closure setpoint (typically 2375 psig), it is sent a signal to close. A simultaneous signal is also sent to the control room indicating to the operator that a close signal was sent to the PORV. PORV closure can be verified by an acoustic monitor installed on the tailpipe downstream of the PORV on all PWRs after the TMI-2 accident. At Davis-Besse, the PORV closure is indicated by a light located on a wall several feet from the operator's control panel. This was available to the operator at Davis-Besse to verify whether the PORV was closed, but was not looked at.
Additionally, there is the SPDS, also a post-TMI improvement, that displays a summary of the most safety significant plant status information on a TV screen.
Both channels were inoperable prior to the event.888 This left the operators with only the pressurizer pressure indicator as a source of determining if the PORV was open or closed. Since the indicator appeared steady, the operator assumed that the PORV had closed, but closed the block valve as a precautionary measure. In actuality, however, the PORV had not closed until some time later into the event.
There have been several stuck open PORVs documented due to a variety of malfunc-tions some of which were identified to be mechanical failure, broken solenoid linkage, inoperability due to corrosion buildup, and sticking caused by foreign material.sse As a precaution, the PORV block valve can be closed to insure no LOCA, but this can only be achieved if the operator closes the block valve by remote-manual operation from the control room. In the Davis-Besse event, the operator did close the block valve to prevent a further decrease in pressure and loss of primary coolant through the PORV when it did not reseat.
0)
Q Possible Solution Knowing that a stuck-open PORV may result in a potentially dangerous scenario (i.e., LOCA), this issue addresses the concern of whether there is a need for an automatic block valve closure in plants that have PORVs.
Considering available control room indicators such as an acoustic monitor, a
~
reliable SPDS and the operator's acute sensitivity 'to the PORV's status because of historical events such as THI-2 and Davis-Besse, another redundant feature (i.e., automating the block valve) would not necessarily result in a significant decrease in core-melt frequency. The acoustic monitor was available to the operator at Davis-Besse; the SPDS was not. However, there is an NRC requirement for the installation of "a concise display of critical plant variables to the control room operators to aid them in rapidly and reliably determining the safety status of the plant."a7s Additionally, there is a DHFT program underway "to determine the need for and, if necessary, the scope of the NRC's SPDS post-implementation reviews."900 The information obtained will " allow an assessment of how well the SPDS objectives are being met and provide the basis for an NRC regulatory position on SPDS post-implementation reviews. Following completion of this program DHFT will, if necessary, work with industry to develop appropriate standards for SPDS availability."900 The staff performed SARs on the three vendor group responses (CE, B&W, W) to TMI Action Plan Item II.K.3(2), " Report on Overall Safety Effect of Power-Operated Relief Valve (PORV) Isolation System." (References 897, 898, and 899).
4 06/30/86 3.125-3 NUREG-0933
(
_ _ _ , . , . _ _ . , _- ,_9.. _
_y.y_ _ - , _ , _ _ . _ . _ _ . _ _ _ , _ ,
The SARs included an estimate of core-melt frequency due to a stuck open PORV-induced SBLOCA. The calculations were based on PORV operating data from April 1, 1980 to March 31, 1983 and concluded that post-TMI actions such as lowering the setpoint of the high pressure reactor trip and raising the setpoint of the PORV opening, eliminating the turbine runback feature, and improving operator capability decreased the challenge to the PORV and the probability of a SBLOCA-PORV sufficiently so as not to warrant a requirement for automatic block valve closure.
The Davis-Besse event may be viewed as another " data point" that should be con-sidered in this determination. However, upon con. sideration of the occurrence of a PORV actuation and the conservative estimates made in the staff's SARs (References 897, 898, and 899), we conclude that the SBLOCA-PORV frequency would still remain within the range of the SBLOCA frequencies given in WASH-140016 (10 2 to 10 4/RY). The opening of the PORV resulted from a loss of all feedwater to the steam generators and is regarded as a legitimate response and fulfillment of the real purpose for incorporating a PORV into the design. Therefore, the Davis-Besse event does not change the statistics for necessary challenge to the PORV. Consequently, the staff's SARs (References 897, 898 and 899) which con-cluded that block valve automation is unnecessary are unaffected.
Also it is clear that the automation of the block valve might reduce the ini-tiator (SBLOCA-PORV) frequency, but not necessarily the net core-melt frequency.
Since it has the potential for spurious actuation (e.g., spurious electrical signal sensed by the block valve could force it closed during a transient re-quiring use of the PORV) which would increase core-melt frequency.
The occurrence at Davis-Besse was the result of an initiator already considered in the SARs, i.e., the failure of the AFW system. It was an occurrence that would have resulted in no other outcome should an automatic block valve have been available because the operator closed the block valve himself as a result of his sensitivity to the PORV from post-TMI training.
CONCLUSION In light of the control room indications available to the operators and the results of the staff SARs (References 897, 898 and 899) that concluded that an automatic PORV isolation system is not necessary, the safety concerns of this issue have been resolved. Thus, this issue should be DROPPED as a new issue.
ITEM 125.I.2.D: CAPABILITY OF THE PORV TO SUPPORT FEED-AND-BLEED DESCRIPTION Historical Background This issue was identified in the EDO memorandum 895 and was also raised at an ACRS Subcommittee meeting on Emergency Core Cooling Systems held on July 31, 1985.
Safety Significance Upon loss of the main and auxiliary feedwater systems, the feedwater flow to the steam generators is insufficient to maintain level. As the level of water in 06/30/86 3.125-4 NUREG-0933
O the steam generators decreases, the average temperature of the RCS increases because of the reduced heat transfer from the primary to the secondary coolant systems. When all steam generators are " dry," the plant emergency procedure requires the initiation of makeup /high pressure injection (MU/HPI) cooling of the primary system. ass This method of decay heat removal is known as " feed-and-bleed" or " bleed-and-feed" depending on the HPI capability of the injection pumps and system design. When this method is initiated, the PORV and high point vents on the RCS, specifically the pressurizer, are locked open breaching one of the plant's radiological barriers and releasing radioactive coolant inside the containment building.sss MU/HPI is often considered a drastic action because of the radioactive contamination of the containment. Nevertheless, MU/HPI cool-ing provides a. diverse method of core cooling if the main and auxiliary feedwater systems should fail.
This issue is based on an ACRS concern that the PORVs are not qualified for the
" hostile" environment in which they are placed when used for feed-and-bleed operation. There are several reasons for this concern. PORVs are usually called upon to respond when all other methods of removing decay heat are not available.
The temperature, pressure, and moisture conditions of the containment environment can create a differentiai thermal expansion of the valve disc and body and may s
cause the PORV to stick,386 failing open or closed, or the PORV can close shortly after beginning feed-and-bleed because of short circuits.
CONCLUSION Under USI A-45, " Shutdown Decay Heat Removal Requirements," the NRC staff is O investigating alternative means of decay heat removal in PWR plants using
( existing equipment or devising new methods. The use of the " feed-and-bleed" procedure is included in this program as well as the need for environmental qualification of the PORV for this method of emergency decay heat removal.
Therefore, this issue is covered in USI A-45.896 REFERENCES
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w./
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897. Memorandum for G. Lainas from F. Rowsome, " Safety Evaluation of the CE Licensees' Responses to TMI Actinn Item II.K.3.2," August 26, 1983.
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O O
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139. SAI-78-649 WA, "A Quantitative Approach for Establishing Limiting Conditions for Operation for ECCS/ECI Components in Commercial Nuclear Power Plants,"
Scientific Applications Inc., March 1979.
140 " Draft Summary Report on a Risk Based Categorization of NRC Technical and Generic Issues," Probabilistic Analysis Staff, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission.
( 141. Regulatory Guide 1.150, " Ultrasonic Testing of Reactor Vessel Welds during Pre-Service and In-Service Examination," U.S. Nuclear Regulatory Commission, June 1981.
06/30/86 R-9 NUREG-0933
R:visi:n 3 142. NRC Letter to Alabama Power Company, " Containment Purging during Normal Plant Operation," (Docket No. 50-348), November 28, 1978.
143. NRC Letter to Nebraska Public Power District, " Containment Purging and Venting during Normal Operation," (Docket No. 50-298), October 22, 1979.
144. Regulatory Guide 1.75, " Physical Independence of Electric Systems,"
September 1978.
145. NRC Interim Criteria for Evaluating Steel Containment Buckling, June 21, 1982.
146. Draft Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," U.S. Nuclear Regulatory Commission, September 1977.
147. Regulatory Guide 1.20, " Comprehensive Vibration Assessment Program for Reactor Internals during Preoperational and Initial Startup Testing,"
U.S. Nuclear Regulatory Commission, May 1976.
148. " Memorandum of Agreement between the Institute of Nuclear Power Operations and the U.S. Nuclear Regulatory Commission," Rev. 1, April 1, 1982.
349. Memorandum for J. Funches from R. Mattson, " Comments on Prioritization of Licensing Improvement Issues," February 2, 1983.
150. Regulatory Guide 1.47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems," U.S. Nuclear Regulatory Commission, May 1973.
151. SECY-82-111, " Requirements for Emergency Response Capability," March 11, 1982.
152. NUREG/CR-2417, " Identification and Analysis of Human Errors Underlying Pump and Valve Related Events Reported by Nuclear Power Plant Licensees,"
U.S. Nuclear Regulatory Commission, February 1982.
153. " Safety System Status Monitoring: Draft Report on Current Industry Practice," Pacific Northwest Laboratories, June 1982.
154. NRC Letter to Construction Permit Holders of B&W Designed Facilities, October 25, 1979.
155. NUREG-0667, " Transient Response of Babcock & Wilcox Designed Reactors,"
U.S. Nuclear Regulatory Commission, May 1980.
156. Memorandum for H. Denton from D. Eisenhut, "NUREG-0667, Transient Response of Babcock & Wilcox Designed Reactors, Implementation Plan," June 3, 1981.
157. Memorandum for D. Eisenhut from G. Lainas, " Status Report on Implementation of NUREG-0667 Category A Recommendations," December 15, 1981.
158. Memorandum for H. Denton from R. Mattson, " Review of Final Report of the B&W Reactor Transient Response Task Force (NUREG-0667)," August 8, 1980.
06/30/86 R-10 NUREG-0333
Revision 3 ,
159.. Memorandum for S. Hanauer from R. Mattson, " Design Sensitivity of B&W y/ Reactors, Item II.E.5.1 of NUREG-0660,". February 26, 1982.
160. Memorandum for R. Mattson from S. Hanauer, " Design Sensitivity of B&W Reactors," June 21, 1982.
161. NUREG/CR-1250, "Three Mile Island: A Report to the Commission and to the Public," U.S. Nuclear Regulatory Commission, January 1980.
162. NRC Letter to All Light Water Reactors, " Containment Purging and Venting During Normal Operation - Guidelines for Valve Operability," September 27, 1979.
163. NUREG-0305, " Technical Report on DC Power Supplies in Nuclear Power Plants," U.S. Nuclear Regulatory Commission, July 1977.
i 164. NUREG-0666, "A Probabilistic Safety Analysis of DC Power Supply Require-ments for Nuclear Power Plants," U.S. Nuclear Regulatory Commission, April 1981.
165. Memorandum for M. Srinivasan from O. Parr and B. Sheron, " Generic Issue (GI) A-30, Adequacy of Safety Related DC Power Supplies, Development of Licensing Guidelines," March 12, 1982.
166. Memorandum for E. Case from R. Mattson, " Task No. D-3 Control Rod Drop Accident (BWRs)," March 6, 1978.
c ./ 167. WCAP-8555, "LOTIC: Long-Term Ice Condenser Containment Code," Westing-house Electric Corporation, April 1976.
168. NRC Letter to Arkansas Power & Light Company, " Order for Modification of License Concerning Primary Coolant System Pressure Isolation Valves,"
(Docket No. 50-313), April 20, 1981.
169. " Report on Standards and Requirements for Electrical Penetration Assemblies for Nuclear Reactor Containment Structures," Oak Ridge National Laboratory, December 13, 1978.
170. NUREG/CR-1345, " Nuclear Power Plant Design Concepts for Sabotage Protec-tion," U.S. Nuclear Regulatory Commission, 1981.
171. Bulletin of the Atomic Scientists, Vol. 32, No. 8, pp. 29-36, " Nuclear Sabotage," M. Flood, October 1976.
172. Federal Register Notices: 43 FR 10370, March 13, 1978; 44 FR 68307, November 28, 1979; 45 FR 37011, May 30, 1980; 46 FR 11666, February 10, 1981.
173. NUREG-0586, "Draf t Generic Environmental Impact Statement on Decommis-sioning of Nuclear Facilities," U.S. Nuclear Regulatory Commission, January 1981.
I O 1 174. NUREG-0585, "TMI Lessons Learned Task Force Final Report," U.S. Nuclear Regulatory Commission, October 1979.
06/30/86 R-11 NUREG-0933
Rsvision 3 175. ZAR-791030-01, " Report of the President's Commission on the Accident at Three Mile Island," J. G. Kemeny et al., November 30, 1979.
176. Memorandum for J. Ahearne f rom M. Carbon, " Comments on the Pause in Licensing," December 11, 1979.
177. Memorandum for N. Moseley from J. Allan, " Operations Team Recommendat. ions
- IE/TMI Unit 2 Investigation," October 16, 1979.
178. EPRI NP-801, "ATWS: A Reappraisal, Part III, Frequency of Anticipated Transients," Electric Power Research Institute, July 1978.
179. NUREG-0020, " Licensed Operating Reactors, Status Summary Report," U.S.
Nuclear Eegulatory Commission, February 1982.
180. NUREG-0580, " Regulatory Licensing Status Summary Report," U.S. Nuclear Regulatory Commission, June 1982.
181. SECY-82-155, "Public Law 96-295, Section 307(B), Study of the Feasibility and Value of Licensing of Nuclear Plant Managers and Senior Licensee Offi-cers," April 12, 1982.
182. NUREG/CR-0672, " Technology, Safety, and Costs of Decommissioning a Reference Boiling Water Reactor Power Station," U.S. Nuclear Regulatory Commission, June 1980.
183. NUREG-0153, " Staff Discussion of 12 Additional Technical Issues Raised by Responses to the November 3, 1976 Memorandum from Director, NRR to NRR Staff," December 1976.
184. Memorandum for R. Vollmer from D. Eisenhet, " Transmittal of Report on Threaded Fastener Experience in Nuclear Power Plants," August 25, 1982.
185. Memorandum for H. Denton from C. Michelson, "AE0D Report on the St. Lucie Natural Circulation Cooldown on June 11, 1980," December 24, 1980.
186. NUREG-0510, " Identification of Unresolved Safety Issues Relating to Nuclear Power Plants," U.S. Nuclear Regulatory Commission, January 1979.
187. NUREG/CR-2300, "PRA Procedures Guide," U.S. Nuclear Regulatory Commission, September 1981.
188. NUREG/CR-2644, "An Assessment of Offsite, Real-Time Dose Measurements for Emergency Situations," U.S. Nuclear Regulatory Commission, April 1982.
189. Memorandum for K. Goller from R. Mattson, " Proposed Changes to Regulatory Guide 1.97," July 29, 1982.
190. Memorandum for W. Dircks from S. Chilk, "Staf f Requirements - Af firmative Session, 11:50 a.m., Friday July 16, 1982," July 20, 1982.
191. NUREG-0799, " Draft Criteria for Preparation of Emergency Operating Procedures," U.S. Nuclear Regulatory Commission, July 1, 1982.
06/30/86 R-12 NUREG-0933 l
Revision 3
(^T ,1 192. NUREG-0899, " Guidelines for Preparation of Emergency Operating Procedures -
Resolution of Comments on NUREG-0799," U.S. Nuclear Regulatory Commission,
' September 3, 1982.
193. Memorandum for J. Martin et al. , from L. Shao, " Division Review Request:
faendments-to 10 CFR Parts 30, 40, 50, 70, and 72 on Decommissioning Criteria for Nuclear Facilities," July 7, 1982.
194. IEEE 500-1977, " Guide to the Collection and Presentation of Electrical, Electronic, and Sensing Component Reliability Data for Nuclear Power Generating Stations," Institute of Electrical and Electronics Engineers.
195. Memorandum for E. Adensam from R. Riggs, " Status on Reactor Coolant Pump Seal Degradation Review," December 9, 1980.
196. Memorandum for H. Denton from S. Hanauer, " Preliminary Ranking of NRR Generic Safety Issues," March 26, 1982.
197. Code of Federal Regulations, Title 10, Energy.
198. NUREG-0698, "NRC Plans for Cleanup Operations at Three Mile Island Unit 2,"
U.S. Nuclear Regulatory Commission, July 1980.
199. NUREG-0683, " Final Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Wastes Resulting from the March 28, 1979 Accident at Three Mile Island Nuclear Station, Unit 2,"
[ ) U.S. Nuclear Regulatory Commission, March 1981.
200. IEEE 603-1977, " Trial-Use Standard Criteria for Safety Systems for Nuclear Power Generating Stations (ANSI N31.30)," Institute of Electrical and Electronics Engineers.
201. NUREG-0460, " Anticipated Transients Without Scram for Light Water Reactors,"
U.S. Nuclear Regulatory Commission, April 1978.
202. Memorandum for G. Cunningham et al. , from K. Goller, " Proposed Amendment to Part 50 on Radiation Programs, Including ALARA," September 10, 1982.
203. SECY-82-157A, " Status Report on the NRR Investigation of the Effects of Electromagnetic Pulse (EMP) on Nuclear Power Plants," July 16, 1982.
204. NUREG-0855, " Health Physics Appraisal Program," U.S. Nuclear Regulatory Commission, March 1982.
205. NUREG-0761, " Radiation Protection Plans for Nuclear Power Reactor Licensees," U.S. Nuclear Regulatory Commission, March 1981.
206. Memorandum for L. Rubenstein from M. Ernst, " Proposed Position Regarding Containment Purge / Vent Systems," April 17, 1981.
207. Nuclear Plant Reliability Data System, Southwest Research Institute.
m 06/30/86 R-13 NUREG-0933
R:. vision 3 208. Regulatory Guide 1.52, " Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," U.S.
Nuclear Regulatory Commission, March 1978.
209. Regulatory Guide 1.140, " Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Wate--Cooled Nuclear Power Plants," U.S. Nuclear Regulatory Commis-sion.
210. NUREG-0885, Issue 2, "U.S. Nuclear Regulatory Commission Policy and Planning Guidance," U.S. Nuclear Regulatory Commission, January 1983.
211. Federal Register Notice 46FR764, "NRC Policy Statement on Cleanup of the Three Mile Island Plant," May 1,1981.
212. NUREG-0772, " Technical Bases for Estimating Fission Product Behavior During LWR Accidents," U.S. Nuclear Regulatory Commission, June 1981.
213. Regulatory Guide 1.3, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Boiling Water Reactors," U.S. Nuclear Regulatory Commission, June 1974.
214. Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors," U.S. Nuclear Regulatory Commission, June 1974.
215. Memorandum for E. Sullivan from R. Bosnak, " Generic Issues," September 17, 1982.
216. Regulatory Guide 1.~108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," U.S. Nuclear Regulatory Commission, August 1977.
217. NUREG/CR-0660, " Enhancement of On-site Emergency Diesel Generator Reli-ability," U.S. Nuclear Regulatory Commission, February 1979.
218. Memorandum for D. Eisenhut, et al., from S. Hanauer, " Diesel Generator Reliability at Operating Plants," May 6, 1982.
219. Memorandum for S. Hanauer from R. Mattson, " Request for Prioritization of BWR Main Staam Line Isolation Valve Leakage as a Generic Issue," July 30, 1982.
220. IE Bulletin No. 82-23, " Main Steam Isolation Valve (MSIV) Leakage," U.S.
Nuclear Regulatory Commission, July 16, 1982.
221. Regulatory Guide 1.48, " Design Limits and Loading Combinations for Seismic Category 1 Fluid System Components," U.S. Nuclear Regulatory Commission, May 1973.
222. NUREG-0479, " Report on BWR Control Rod Drive Mechanical Failures," U.S.
Nuclear Regulatory Commission, January 1979.
06/30/86 R-14 NUREG-0933
1 I
Revision 3 (A
(s) 223. NUREG-0462, " Technical Report on Operating Experience with BWR Pressure Relief Valves," U.S. Nuclear Regulatory Commission, July 1978.
224. NUREG-0654, " Criteria for Preparation and Evaluation of Radiological
. Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," U.S. Nuclear Regulatory Commission, November 1980.
225. Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation)," U.S. Nuclear Regulatory Commission, February 1978.
226. Regulatory Guide 1.8, " Personnel Selection and Training," U.S. Nuclear Regulatory Commission, May 1977.
227. NUREG/CR-0130, " Technology, Safety, and Costs of Deconmissioning a Reference Pressurized Water Reactor Power Station," U.S. Nuclear Regula-tory Commission, June 1978.
228. SECY-81-450, " Development of a Selective Absorption System Emergency -
Unit," July 27, 1981.
229. Memorandum for T. Speis from R. Houston, " Containment Venting and Purging - Completion of TMI Action Plan Item II.E.4.4(4)," March 3, 1982.
230. Memorandum for R. Mattson from T. Speis, ". Containment Purge and Venting -
Completion of TMI Action Plan Item II.E.4.4(5)," April 9, 1982.
)
231. Memorandum for W. Dircks from R. Mattson, " Status Report on Containment Purge Evaluations," June 2, 1982.
R 232. SECY-81-1688, " Response to Commission Request for Information on Financial Considerations in Licensing Proceedings," July 13, 1981.
233. IEEE P-827, "A Method for Determining Requirements for lastrumentation Control and Electrical Systems and Equipment Important to Safety,"
Institute of Electrical and Electronics Engineers.
234. Federal Register, Vol. 47, No. 46, "10 CFR Part 2, General Statement of Policy and Procedure for Enforcement Actions," March 9, 1982.
235. Memorandum for H. Denton from R. DeYoung, "TMI Action Plan Items Still Pending," June 10, 1982.
236. Memorandum for W. Dircks from R. DeYoung, "TMI Action Plan - Completed
- Items," June 30, 1982.
237. SECY-80-366, "NRC Legislative Program for 97th Congress," August 6, 1980.
238. NRC-INP0 Memorandum of Agreement, June 1, 1981.
e r 9 9b.
240. SECY-81-153, " Nuclear Data Link," March 11, 1981.
06/30/86 R-15 NUREG-0933
R@ vision 3 241. NUREG/CR-1440, " Light Water Reactor Status Monitoring During Accident Conditions," U.S. Nuclear Regulatory Commission, May 1980.
242. NUREG/CR-2100, " Boiling Water Reactor Status Monitoring During Accident Conditions," U.S. Nuclear Regulatory Commission, May 1981.
243. NUREG/CR-2278, " Light Water Reactor Engineered Safety Features Status Monitoring," U.S. Nuclear Regulatory Commisson, October 1981.
244. NUREG/CR-2147, " Nuclear Control Room Annunciators," U.S. Nuclear Regulatory Commission, October 1981.
245. RIL-124, " Control Room Alarms and Annunciators," U.S. Nuclear Regulatory Commission.
246. RIL-98, " Light Water Reactor Status Monitoring During Accident Conditions,"
U.S. Nuclear Regulatory Commission, August 18, 1980.
247. Inspection and Enforcement Manual, Office of Inspection and Enforcement, U.S. Nuclear Regulatory Commission.
248. Memorandum for W. Dircks fro n R. DeYoung, "TMI Action Plan - Completed Items," December 28, 1981.
249. " Branch Technical Position on Waste Form," Office of Nuclear Materials Safety and Safeguards, U.S. Nuclear Regulatory Commission, February 14, 1983.
250. SECY-81-440, " Nuclear Power Plant Staff Working Hours," July 22, 1981.
251. SECY-79-330E, " Qualifications of Reactor Operators," July 30, 1979.
252. NRR-80-117, " Study of Requirements for Operator Licensing," February 4, 1980.
253. ANSI /ANS 3.1, " Selection, Qualification, and Training of Personnel for Nuclear Power Plants," American National Standards Institute, 1981.
254. Memorandum to N. Palladino from M. Udall, Chairman, Committee on Interior and Insular Affairs, U.S. House of Representatives, June 4, 1982.
255. Memorandum to M. Udall, Chairman, Committee on Interior and Insular Affairs, U.S. House of Representatives, from N. Palladino, June 30, 1982.
256. Memorandum for W. Dircks from R. DeYoung, "TMI Action Plan - Completed Items," June 2, 1982.
257. NUREG-0728, " Report to Congress - NRC Incident Respense Plan," U.S.
Nuclear Regulatory Commission, September 1980.
258. NUREG-0845, " Agency Procedure for the NRC Incident Response Plan," U.S.
Nuclear Regulatory Commission, March 1982.
259. Memorandum for J. Snlezek from J. Taylor, "TMI Action Plan item II.J.1.2, Modification of Vendor Inspection Program," October 13, 1982.
06/30/86 R-16 NUREG-0933
Ravision 3 260. SECY-81-494, " Integrated Operational Experience Reporting System,"
g August 18, 1981.
261. Federal Register Notice 46FR53594, "NRC Regulatory Agenda," October 29, 1981.
262. BNL/NUREG-28955, "PWR Training Simulator and Evaluation of the Thermal-Hydraulic Models for Its Main Steam Supply System," Brookhaven National Laboratory, 1981.
263. BNL/NUREG-29815, "BWR Training Simulator and Evaluation of the Thermal-Hydraulic Models for Its Main Steam Supply System," Brookhaven National Laboratory, 1981.
264. BNL/NUREG-30602, "A PWR Training Simulator Comparison with RETRAN for a Reactor Trip from Full Power," Brookhaven National Laboratory,1981.
265. Memorandum for the Commissioners from W. Dircks, " Enforcement Policy,"
March 18, 1980.
266. SECY-80-139A, "NRC Enforcement Program," August 27, 1980.
267. Memorandum for R. Purple from R. Minogue, "TMI Action Plan," October 24, 1980.
268. Memorandum for W. Dircks from V. Stello, " Assignment of Resident (9 Inspectors to Nuclear Steam System Suppliers and Architect-Engineers,"
Q 1 September 14,~1981.
269. IE Circular No. 80-15, " Loss of Reactor Coolant Pump Cooling and Natural Circulation Cooldown,"'U.S. Nuclear Regulatory Commission, June 20, 1980.
270. Memorandum for C. Michelson from H. Denton, " Report on St. Lucie Natural Circulation Cooldown," April 6, 1981.
271. INP0/NSAC Significant Operating Experience Report 81-17, " Potential for Steam Line Rupture to Affect Auxiliary Feedwater System," November 11, 1981.
i 272. Memorandum for J. Gagliardo from D. Eisenhut, " Potential Failure of Turbine Driven Auxiliary Feedwater Pump Steam Supply Line - Fort Calhoun,"
October 8, 1982.
273. Memorandum for H. Denton from C. Michelson, " Technical Review Report, Postulated Loss of Auxiliary Feedwater System Resulting from Turbine Driven Auxiliary Feedwater Pump Steam Supply Line Rupture," February 16, 1983. ,
t 274. Letter to G. Knighton (NRC) from K. Baskin (Southern California Edison Company), " Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station Units 2 and 3," October 29, 1982.
275. NUREG/CR-1614, " Approaches to Acceptable Risk: A Critical Guide," U.S.
Nuclear Regulatory Commission, September 1980.
06/30/86 R-17 NUREG-0933 s*
a_m -#..__._----___-+__ c,_g, p, ____g..my -,._wy-_-y__ _. , s.,_y .--- _.3g. mp,,,y9w%.,,y7,,g,gg. .,, , ,,_,,, _,,,,c- . , , . , _ _ , , , , , . - ,
R3 vision 3 276. NUREG/CR-1539, "A Methodology and a Preliminary Data Base for Examining the Health Risks of Electricity Generation from Uranium and Coal Fuels,"
U.S. Nuclear Regulatory Commission, August 1980.
277. NUREG/CR-1930, "Index of Risk Exposure and Risk Acceptance Criteria,"
U.S. Nuclear Regulatory Commission, February 1981.
278. NUREG/CR-1916, "A Risk Comparison," U.S. Nuclear Regulatory Commission, February 1981.
279. NUREG/CR-2040, "A' Study of the Implications of Applying Quantitative Risk Criteria in the Licensing of Nuclear Power Plants in the U.S.," U.S.
Nuclear Regulatory Commission, March 1981.
280. SECY-80-331, "NRC Training Program," July 14, 1980.
281. Memorandum for H. Denton et. al. , f rom C. Michelson, " Case Study Report -
Failure of Class 1E Safety-Related Switchgear Circuit Breakers to Close on Demand," August 4, 1982.
282. Memorandum for C. Michelson from H. Denton, "AE0D Preliminary Report on Failures of Class 1E Safety-Related Switchgear Circuit Breakers to Close on Demand," September 23, 1982.
283. Executive Order 12379, " Termination of Boards, Committees, and Commissions," August 17, 1982.
284. Letter to N. Palladino, Chairman, NRC, from G. Keyworth, Director, OSTP, July 21, 1982.
285. Letter to G. Keyworth, Director, OSTP, from N. Palladino, Chairman, NRC, July 23, 1982.
286. Letter to T. Pestorius, Chairman, Committee for Interagency Radiation Policy Coordination, OSTP, from R. Minogua, NRC, August 27, 1982.
287. SECY-81-600A, " Revised General Statement of Policy and Procedure for Enforcement Actions," December 14, 1981.
288. NED0-10174, Revision 1, " Consequences of a Postulated Flow Blockage Incident in a Boiling Water Reactor," General Electric Company, October 1977.
289. NUREG/CR-2075 " Standards for Psychological Assessment of Nuclear Facility Personnel," U.S. Nuclear Regulatory Commission, July 1981.
290. NUREG/CR-2076, " Behavioral Reliability Program for the Nuclear Industry,"
U.S. Nuclear Regulatory Commission, July 1981.
291. Memorandum for E. Jordan et. al., from R. Bernero," Proposed Rule Review Request-10 CFR Part 21: Reporting of Defects and Noncompliance,"
September 28, 1982.
06/30/86 " 18 NUREG-0933
'Ravision 3 O 292. Memorandum for R. Minogue from R. DeYoung, " Proposed Rule Amending 10 CFR (j Parts 50.55(e) and 21: RES Task Numbers RA 128-1 and RA 808-1," July 13, 1982.
293. Federal Register Notice 47 FR 18508, "NRC Regulatory Agenda," April 29, 1982.
294. Federal Register Notice 47 FR 48960, "NRC Regulatory Agenda," October 28, 1982.
295. BNL-NUREG-31940, " Postulated SRV Line Break in the Wetwell Airspace of Mark I and Mark II Containments - A Risk Assessment," Brookhaven National Laboratory, October 1982.
296. " Development of the Automated Vendor Selection System," Gasser Associates, Inc., June 30, 1980.
297. Memorandum for W. Dircks from R. DeYoung, "TMI Action Plan-Completed Item," October 29, 1982.
298. Memorandum for W. Dircks from V. Stello, "TMI Action Plan-Status Report,"
April 17, 1981.
299. NUREG/CR-1482, " Nuclear Power Plant Simulators: Their Use in Operator Training and Requalification," U.S. Nuclear Regulatory Commission, August 1980.
300. NUREG/CR-2353, " Specification and Verification of Nuclear Power Plant Training Simulator Response Characteristics," U.S. Nuclear Regulatory Commission, 1982.
301. Memorandum for R. Emrit from P. Goldman, " Draft Report on the Prioritiza-tion of Non-NRR TMI Action Plan Items," December 29, 1982.
302. Memorandum for H. Denton from C. Michelson, " Operational Restrictions for Class IE 120 VAC Vital Instrument Buses," July 15, 1980.
303. Memorandum for C. Michelson from H. Denton, "LCO for Class IE Vital Instru-ment Buses in Operating Reactors," September 29, 1980.
304. UCID-19469, "120 VAC Vital Instrument Buses and Inverter Technical Specifi-cations," Lawrence Livermore National Laboratory, October 28, 1982.
305. Memorandum to Distribution from J. Davis, "NMSS Procedure for Review of Routine Inspection Operational Data and Licensee Event Reports," March 9, 1982.
306. IEEE Catalog No. TH0073-7, " Record of the Working Conference on Advanced Electrotechnology Applications to Nuclear Power Plants, January 15-17, 1980, Washington, D.C.," Institute of Electrical and Electronics Engineers.
307. EPRI NP-2230, "ATWS: A Reappraisal, Part 3," Electric Power Research O Institute, 1982.
06/30/86 R-19 NUREG-0933
R:visicn 3 308. SECY-82-352, " Assurance of Quality," August 20, 1982.
309. SECY-82-1, " Severe Accident Rulemaking and Related Matters," January 4, 1982.
310. Memorandum for W. Dircks from S. Chilk, " Staff Requirements - Briefing on Status and Plan for Severe Accident Rulemaking (SECY-82-1)," January 29, 1982.
311. SECY-82-1A, " Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation," July 16, 1982.
312. NUREG/CR-0165, "A Value-Impact Assessment of Alternate Containment Concepts," U.S. Nuclear Regulatory Commission, June 1978.
313. NUREG/CR-2063, " Effects of the Accident of Three Mile Island on Property Values and Sales," U.S. Nuclear Regulatory Commission, March 1981.
314. NUREG/CR-2749, Volume 12, " Socioeconomic Impacts of Nuclear Generating Stations - Three Mile Island Case Study," U.S. Nuclear Regulatory Commission, July 1982.
315. Memorandum of Understanding Between the Federal Emergency Management Agency and the Nuclear Regulatory Commission, " Incident Response,"
October 22, 1980.
316. Memorandum of Understanding Between the Nuclear Regulatory Commission and the Federal Emergency Management Agency, " Radiological Emergency Planning and Preparedness," November 4, 1980.
317. Memorandum for G. Lainas from F. Miraglia, "CRGR Package for MPA B-71, 120 VAC Vital Instrument Buses and Inverter Technical Specifications,"
November 23, 1982.
318. Memorandum for H. Denton et. al. , from C. Michelson, "An Analysis of the Abnormal Transient Operating Guidelines (ATOG) as Applied to the April 1981 Overfill Event at Arkansas Nuclear One - Unit 1," April 9, 1982.
319. Memorandum for C. Michelson from D. Eisenhut, " Review of Abnormal Tran-sient Operating Guidelines (ATOG) as Applied to the April 8, 1981 Overfill Event at Arkansas Nuclear One - Unit 1," July 30, 1982.
320. Memorandum for C. Michelson f rom H. Denton, " Review of the Case Study of the Abnormal Transient Operating Guidelines (ATOG) as Applied to the April 1981 Overfill Event at Arkansas Nuclear One - Unit 1," October 7, 1982.
321. Memorandum for Commissioner Ahearne from W. Dircks, "AE00 Report on Arkansas Unit 1 Overfill Event," November 1, 1982.
322. AE00/C201, " Report on The Safety Concern Associated with Reactor Vessel Level Instrumentation in Boiling Water Reactors," Office for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, January 1982.
06/30/86 R-20 NUREG-0933
R: vision 3 (S
! ) 323. Memorandum for C. Michelson from H. Denton, "AE00 January 1982 Report on U Safety Concern Associated with Reactor Vessel Level Instrumentation in Boiling Water Reactors," March 19, 1982.
324. NUREG-0785, " Safety Concerns Associated with Pipe Breaks in the BWR Scram System," U.S. Nuclear Regulatory Commission, April 1981.
325. Letter to All BWR Licensees from D. Eisenhut, " Safety Concerns Associated with Pipe Breaks in th'e BWR Scram System," April 10, 1981.
326. NED0-24342, "GE Evaluation in Response to NRC Request Regarding BWR Scram System Pipe Breaks," General Electric Company, April 1981.
327. Letter to D. Eisenhut (NRC) from G. Sherwood (GE), " Safety Concerns Associated with Pipe Breaks in the BWR Scram System," April 30, 1981.
328. NUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping," U.S. Nuclear Regulatory Commission, August 1981.
329. Letter to All GE BWR Licensees (Except Humboldt Bay) from D. Eisenhut,
" Safety Concerns Associated with Pipe Breaks in the BWR Scram System (Generic Letter 81-34)," August 31, 1981.
330. AE00/C003, " Report on Loss of Offsite Power Event at Arkansas Nuclear One, Units 1 and 2, on April 7,1980," Office for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, October 15, 1980.
\'
331. Memorandum for C. Michelson from H. Denton, "NRR Responses to AE0D Recom-mendations on the Arkansas Loss of Offsite Power Event of April 7, 1980,"
February 13, 1981.
332. Letter to All BWR Applicants for cps, Holders of cps, and Applicants for OLs from D. Eisenhut, " Safety Concerns Associated with Pipe Breaks in the BWR Scram System (Generic Letter 81-35)," August 31, 1981.
333. SECY-82-445, " Proposal to Assign Two Resident Inspectors to Each Reactor Construction Site," November 1, 1982.
334. SECY-82-478, " Resident Inspection Program," December 6, 1982.
335. NRC Manual Chapter 0516, U.S. Nuclear Regulatory Commission, March 23, 1982.
336. NUREG-0834, "NRC Licensee Assessments," U.S. Nuclear Regulatory Commission, August 1981.
337. NUREG/CR-2672, "SBLOCA Outside Containment at Browns Ferry Unit One -
Accident Sequence Analysis," U.S. Nuclear Regulatory Commission, November 1982.
338. NUREG/CR-2744, " Human Reliability Data Bank for Nuclear Power Plant Opera-tions," U.S. Nuclear Regulatory Commission, November 1982.
e G]
06/30/86 R-21 NUREG-0933
R. vision 3 339. NUREG/CR-1278, " Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications," U.S. Nuclear Regulatory Commission, February 1983.
340. Memorandum for H. Denton from J. Fouchard, " Draft Report on the Prioritiza-tion of Non-NRR TMI Action Plan Items," January 17, 1983.
341. NUREG/CR-2255, " Expert Estimation of Human Error Probabilities in Nuclear Power Plant Operations: A Review of Probability Assessment and Scaling,"
U.S. Nuclear Regulatory Commission, May 1982.
342. NUREG/CR-2743, " Procedures for Using Expert Judgment to Estimate Human Error Probabilities in Nuclear Power Plant Operations," U.S. Nuclear Regulatory Commission, February 1983.
343. NUREG/CR-2254, " Workbook for Conducting Human Reliability Analysis,"
U.S. Nuclear Regulatory Commission, February 1983.
344. NUREG/CR-1205, " Data Summaries of Licensee Event Reports of Pumps at U.S. Commercial Nuclear Power Plants, January 1, 1972 through April 30, 1978," U.S. Nuclear Regulatory Commission, January 1980.
345. NUREG/CR-1205, Rev. 1, " Data Summaries of Licensee Event Reports of Pumps at U.S. Commercial Nuclear Power Plants, January 1, 1972 to September 30, 1980," U.S. Nuclear Regulatory Commission, January 1982.
346. NUREG/CR-1363, " Data Summaries of Licensee Event Reports of Valves at U.S. Commercial Nuclear Power Plants, January 1, 1976 through December 31, 1978," U.S. Nuclear Regulatory Commission, June 1980.
347. NUREG/CR-1363, Rev. 1, " Data Summaries of Licensee Event Reports of Valves at U.S. Commercial Nuclear Power Plants, January 1, 1976 to December 31, 1980," U.S. Nuclear Regulatory Commission, October 1982.
348. NUREG/CR-1362. " Data Summaries of Licensee Event Reports of Diesel Genera-tors at U.S. Commercial Nuclear Power Plants, January 1, 1976 through December 31, 1978," U.S. Nuclear Regulatory Commission, March 1980.
349. NUREG/CR-1331, " Data Summaries of Licensee Event Reports of Control Rods and Drive Mechanisms at U.S. Commercial Nuclear Power Plants, January 1, 1972 through April 30, 1978," U.S. Nuclear Regulatory Commission, February 1980.
350. NUREG/CR-1730, " Data Summaries of Licensee Event Reports of Primary Containment Penetrations at U.S. Commercial Nuclear Power Plants, January 1, 1972 through December 31, 1978," U.S. Nuclear Regulatory Commission, September 1980.
351. NUREG/CR-1740, " Data Summaries of Licensee Event Reports of Selected Instrumentation and Control Components at U.S. Commercial Nuclear Power Plants from January 1, 1976 to December 31, 1978," U.S. Nuclear Regulatory Commission, May 1981.
06/30/86 R-22 NUREG-0933
R2 vision 3
) 352. Memorandum for C. Michelson from E. Brown, " Internal Appurtenances in V LWRs," December 24, 1980.
353. NUREG/CR-2641, "The In-Plcnt Reliability Data Base for Nuclear Power Plant Components: Data Collection and Methodology Report," U.S. Nuclear Regulatory Commission, July 1982.
354. NUREG/CR-2886, "The In-Plant Reliability Data Base for Nuclear Power Plant Components: Data Report - The Pump Component," U.S. Nuclear Regulatory Commission, January 1983.
355. EGG-EA-5502, " User's Guide to BFR, A Computer Code Based on the Binomial i Failure Rate Common Cause Model," EG&G, Inc., July 1982.
356. EGG-EA-5623, Rev. 1, " Common Cause Fault Rates for Instrumentation and Control Assemblies: Estimates Based on Licensee Event Reports at U.S.
Commercial Nuclear Power Plants, 1976-1978," EG&G, Inc., September 1982. ,
357. EGG-EA-5485, Rev. 1, " Common Cause Fault Rates for Valves: Estimates Based on Licensee Event Reports at U.S. Commercial Nuclear Power Plants, 1976-1980," EG&G, Inc., September 1982.
358. NUREG/CR-2099, Rev.1, " Common Cause Fault Rates for Diesel Generators:
Estimates Based on Licensee Event Reports at U.S. Commercial Nuclear Power
- Plants, 1976-1978," U.S. Nuclear Regulatory Commission, June 1982.
359. NUREG/CR-1401, " Estimators for the Binomial Failure Rate Common Cause b Model," U.S. Nuclear Regulatory Commission, April 1980.
360. EGG-EA-5289, Rev. 1, " Common Cause Fault Rates for Pumps: Estimates Based on Licensee Event Reports at U.S. Commercial Nuclear Power Plants, January 1, 1972 through September 30, 1980," EG&G, Inc., August 1982.
j 361. JBFA-101-82, " Common Cause Screening Methodology Project (FY 81 Technical Progress Report)," JBF Associates, Inc., February 1982.
362. NUREG/CR-2542,'" Sensitivity Study Using the FRANTIC II Code for the Un-availability of a System to the Failure Characteristics of the Components and the Operating Conditions," U.S. Nuclear Regulatory Commission, February 1982.
363. NUREG/CR-2332, " Time Dependent Unavailability of a Continuously Monitored Component," U.S. Nuclear Regulatory Commission, August 1981.
364. Memorandum for W. Dircks from S. Chilk, " Systematic Assessment of Licensee i Performance," October 20, 1981.
365. NUREG/CR-2515, " Crystal River 3 Safety Study," U.S. Nuclear Regulatory Commission, December 1981.
366. NUREG/CR-2787, " Interim Reliability Evaluation Program: Analysis of the Arkansas Nuclear One - Unit One Nuclear Power Plant," U.S. Nuclear Q Regulatory Commission, June 1982.
06/30/86 R-23 NUREG-0933
R: vision 3 367. NUREG/CR-2802, " Interim Reliability Evaluation Program: Analysis of the Browns Ferry Unit 1 Nuclear Plant," U.S. Nuclear Regulatory Commission, July 1982.
368. Memorandum for ACRS Members from C. Michelson, " Failure of a Feedwater Flow Straightener at San Onofre Nuclear Station, Unit 1," June 13, 1979.
369. SECY-82-396A, " Withdrawal of SECY-82-396 (Federal Policy Statement on Use of Potassium Iodide)," October 15, 1982.
370. SECY-81-676, " Delegation of Rulemaking Authority to the EDO," December 3, 1981.
371. SECY-82-187, " Revised Guidelines for Value-Impact Analyses," May 7, 1982.
372. SECY-82-477, " Draft Report of the Regulatory Reform Task Force,"
November 3, 1982.
373. NUREG-0499, " Preliminary Statement on General Policy for Rulemaking to Improve Nuclear Power Plant Licensing," U.S. Nuclear Regulatory Commission, December 1978.
374. Memorandum for J. Hendrie from L. Bickwit, " Review of Commission Delegation of Authority," October 4, 1979.
375. Memorandum for R. Minogue from R. Bernero, " Charter of the Regulatory Analysis Branch," October 9, 1981.
376. NRC Letter to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits, " Supplement 1 to NUREG-0737, Requirements for Emergency Response Capability (Generic Letter No. 82-33),"
December 17, 1982.
377. Memorandum for W. Minners from B. Snyder, " Schedule for Resolving and Completing Generic Issues," December 16, 1982.
378. Memorandum for S. Boyd from M. Srinivasan, "FY 1983-FY 1984 Office of Nuclear Reactor Regulation Operating Plan," November 17, 1982.
379. Memorandum for H. Denton from R. DeYoung, " Draft Report on the Prioritiza-tion of Non-NRR TMI Action Plan Items," January 24, 1983.
380. NED0-10174, " Consequences of a Postulated Flow Blockage Incident in a Boiling Water Reactor," General Electric Company, May 1980.
381. Memorandum for W. Minners from O. Parr, "Prioritization of Proposed Generic Issue on CRD Accumulator Check Valve Leakage," August 13, 1984.
382. Memorandum for W. Minners from R. Mattson, " Schedules for Resolving and Completing Generic Issues," January 21, 1983.
383. Memorandum for W. Dircks from R. Mattson, "Closcout of TMI Action Plan I.C.1(4), Confirmatory Analyses of Selected Transients," November 12, 1982.
06/30/86 R-24 NUREG-0933
Rnvision 3 384. Memorandum for T. Speis from R. Vollmer, " Schedules for Resolving and V Completing Generic Issues," February 1, 1983.
385. Memorandum for T. Murley from R. Mattson, "Use of Equipment Not Classified as Essential to Safety in BWR Transient Analysis," March 10, 1981.
386. Memorandum for T. Novak from R. Frahm, " Summary of Meeting with General Electric on the Use of Non-Safety Grade Equipment," March 7, 1979.
387. NUREG-0410, "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants," U.S. Nuclear Regulatory Commission, January 1978.
388. NUREG-0577, Revision 1, " Potential for Low Fracture Toughness and Lamellar Tearing in PWR Steam Generator and Reactor Coolant Pump Supports," U.S.
Nuclear Regulatory Commission, October 1983.
389. " Indian Point Probabilistic Safety Study," Power Authority of the State of New York and Consolidated Edison Company of New York, Inc., 1932.
390. NUREG-0850, " Preliminary Assessment of Core Melt Accidents at the Zion and Indian Point Nuclear Power Plants and Strategies for Mitigating _Their Effects," U.S. Nuclear Regulatory Commission, November 1981.
391. Memorandum for E. Reeves from J. Knight, " Zion Liquid Pathway Analysis,"
August 8, 1980.
392. Memorandum for J. Funches from R. Mattson, " Request for Approval to Work v on Low Priority Generic Safety Issues," November 5, 1982.
393. "TMI-2 Recovery Program Estimate," Rev. 1, General Public Utilities Corp.,
July 1981.
394. Memorandum for S. Hanauer et al. , from D. Eisenhut, " Operating Reactor Event Memorandum No. 81-31: Loss of Direct Current (DC) Bus at Millstone Unit 2," March 31, 1981.
395. Memorandum for H. Denton from C. Michelson, " Millstone Unit 2 - Reactor Trip Following De-Energization of a 125 V DC Bus," November 5,1981.
396. Memorandum for C. Michelson from H. Denton, "AEOD November 1981 Report on the Millstone Unit 2 Loss of 125 V DC Bus Event," January 4, 1982.
397. IEEE 279-1971, " Criteria for Protection Systems for Nuclear Power Generat-ing Stations (ANSI N42.7-1972)," Institute of Electrical and Electronics Engineers.
398. Memorandum for R. Tedesco from T. Speis, " Identification of Protection System Instrument Sensing Lines," April 29, 1982.
399. Memorandum for T. Speis from R. Tedesco, " Identification of Protection System Instrument Sensing Lines," June 7, 1982.
400. Memorandum for V. Stello from H. Denton, " Standard Review Plan Guidance for Identification of Protection System Instrument Lines," December 29, 1982.
06/30/86 R-25 NUREG-0933
R visi:n 3 401. Memorandum for H. Denton from V. Stello, " Proposed Standard Review Plan Guidance for Identification of Protection System Instrument Lines,"
January 27, 1983.
402. Letter to D. Eisenhut from T. Dente (BWR Owners' Group), " Analysis of Scram Discharge Volume System Piping Integrity, NED0-22209 (Prepublication Form)," August 23, 1982.
403. Letter to K. Eccleston from T. Dente (BWR Owners' Group), " Transmittal of Supporting Information on Application of Scram Time Fraction to Scram Discharge Volume (SDV) Pipe Break Probability as Used in NE00-22209,"
January 28, 1983.
404. Letter to S. Israel from J. Hickman (Sandia Laboratories), " Review and Evaluation of the Indian Point Probabilistic Safety Study," August 25, 1982.
405. Memorandum for W. Minners from A. Thadani et al., " Probability of Core Melt Due to Component Cooling '4ater System Failures," January 19, 1983.
406. Memorandum for W. Dircks from R. DeYoung, "TMI Action Plan - Status Report,"
March 14, 1982.
407. Memorandum for W. Dircks from R. DeYoung, "TMI Action Plan - Completed Item," May 11, 1982.
408. Memorandum for W. Dircks from R. DeYoung, "TMI Action Plan Completed Items,"
January 26, 1983.
409. Memorandum for W. Minners from W. Mills, "Prioritization of Generic Issue III.D.3.5, Radiation Worker Data Base," February 22, 1983.
410. Memorandum for T. Speis from R. Browning, " Draft Report on the Prioritiza-tion of Non-NRR TMI Action Plan Items," April 1, 1983.
411. SLI-8211, " Review of BWR Reactor Vessel Water Level Measurement Systems,"
- 5. Levy, Inc., July 1982.
412. Memorandum for T. Speis from J. Funches, "Prioritization of Generic Issues - Environmental and Licensing Improvements," February 24, 1983.
413. Memorandum for D. Eisenhut from E. Jordan, " Main Steam Isolation Valve (MSIV) Survey," July 1, 1982.
414. Memorandum for W. Minners from L. Hulman, " Consequence Analyses for BWR Main Steam System Leakage Pathway Generic Issue Evaluation," December 9, 1982.
415. Memorandum for W. Minners from L. Hulman, "MSIV Leakage Consequences,"
December 23, 1982.
416. NUREG/CR-1908, " Criteria for Safety-Related Nuclear Power Plant Operator Actions: Initial Pressurized Water Reactor (PWR) Simulator Exercises,"
U.S. Nuclear Regulatory Commission, September 1981.
06/30/86 R-26 NUREG-0933
Revision 3 b) 417. NUREG/CR-2598, " Nuclear Power Plant Control Room Task Analysis: Pilot Study for Pressurized Water Reactors," U.S. Nuclear Regulatory Commission, July 1982.
' 418. NUREG/CR-2534, " Criteria for Safety-Related Nuclear Power Plant Operator Actions: Initial Boiling Water Reactor (BWR) Simulated Exercises,"
l U.S. Nuclear Regulatory Commission, November 1982.
l 419. NUREG/CR-3092, " Criteria for Safety-Related Nuclear Power Plant Operator Actions: Initial Simulator to Field Data Calibration," U.S. Nuclear Regulatory Commission, February 1983.
420. IE Bulletin No. 80-14, " Degradation of BWR Scram Discharge Volume Capabil- '
ity," U.S. Nuclear Regulatory Commission.
421. IE Bulletin No. 80-17, " Failure of 76 of 185 Control Rods to Fully Insert During a Scram at a BWR," U.S. Nuclear Regulatory Commission.
422. NRC Letter to All BWR Licensees, "BWR Scram Discharge System," December 9, 1980.
423. Memorandum for R. Mattson from D. Eisenhut, " Status of Long-Term Followup of the Indian Point Unit 2 Flooding Event," May 13, 1982.
424. Memorandum for F. Schroeder from T. Speis,." Designation of Inadvertent i Containment Flooding as a Generic Issue," August 5, 1982.
V ~
425. " Zion Probabilistic Safety Study," Commonwealth Edison Company, 1981.
426. Memorandum for T. Novak from G. Lainas and V. Noonan, "NRR Input to SER on Indian Point Unit No. 2 Flood in Containment Due to Containment Cooler Service Water Leaks on 10/17/80," April 3, 1981.
427. Memorandum for T. Speis from R. Mattson, "Close-out of TAP-A-16, Steam Effects on BWR Core Spray Distribution (TACS-40066)," March 29, 1983.
428. Memorandum for W. Minners from P. Hayes, " Generic Safety Issue No. 51, Improved Reliability of Open Service Water Systems," April 5, 1983.
429. Memorandum for J. Knight from E. Sullivan, " Review ACRS Consultant Report,"
January 10, 1980.
430. Memorandum for K. Seyfrit from E. Imbro, " Flow Blockage in Essential Raw Cooling Water System Due to Asiatic Clam Intrusion," March 28, 1983.
431. EPRI NP-1138, " Limiting Factor Analysis of High-Availability Nuclear Plants,"
Electric Power Research Institute, September 1979.
432. SECY-82-296, " Resolution of AE0D Combination LOCA Cor.cern," July 13, 1982.
433. Memorandum for C. Michelsen from E. Brown, " Degradation of Internal h
t Appurtenances and/or Loose Parts in LWRs," June 15, 1982.
06/30/86 R-27 NUREG-0933
.. . - _ - . _ _ - . ~ _ .-
R: vision 3 434. Memorandum for H. Denton and V. Stello from C. Michelson, " Flow Blockage in Essential Equipment at ANO Caused by Corbicula sp. (Asiatic Clams),"
October 20, 1980.
435. Letter to N. Palladino, Chairman, U.S. Nuclear Regulatory Commission, from P. Shewmon, Chairman, Advisory Committee on Reactor Safeguards, " Control Room Habitability," August 18, 1982.
436. Letter to J. Ray, Chairman, Advisory Committee on Reactor Safeguards, from W. Dircks, " August 18, 1982 ACRS Letter on Control Room Habitability,"
January 31, 1983.
437. Memorandum for H. Denton from R. Minogue, " Draft Report on the Prioritiza-tion of Non-NRR TMI Action Plan Items," March 29, 1983.
438. Memorandom for G. Cunningham et. al., from W. Dircks, "NRC Actions Required by Enactment of the Nuclear Waste Policy Act of 1982," January 19, 1983.
439. Regulatory Guide 1.149, " Nuclear Power Plant Simulators for Use in Operator Training," U.S. Nuclear Regulatory Commission, April 1981.
440. Memorandum for W. Minners from D. Ziemann, " Schedules for Resolving and Completing Generic Issues," April 5, 1983.
441. Memorandum for H. Denton from R. DeYoung, " Commission Paper on the Priori-tization of Generic Safety Issues," April 20, 1983.
442. Memorandum for R. Emrit from T. Rothschild, " Establishing Priorities for Generic Safety Issues," April 21, 1983.
443. Memorandum for W. Dircks from R. Mattson, " Closeout of NUREG-0660 Item II.E.5.1, Design Sensitivity of B&W Plants for Operating Plants,"
March 15, 1983.
444. NRC Letter to Pubic Service Electric and Gas Company " Meeting Summary -
Salem Unit 1 Failure of Reactor Trip Breakers," Docket No. 50-272, March 14, 1983.
445. NUREG-1000, Volume 1, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant," U.S. Nuclear Regulatory Commission, April 1983.
NUREG-1000, Volume 2, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant," U.S. Nuclear Regulatory Commission, August 1983.
446. Memorandum for Chairman Ahearne from C. Michelson, "New Unresolved Safety Issues," August 4, 1980.
447. NUREG-0977, "NRC Fact Finding Task Force Report on the ATWS Events at Salem Nuclear Generating Station Unit 1 on February 22 and 25, 1983,"
U.S. Nuclear Regulatory Commission, March 1983.
1 448. Memorandum for F. Rowsome from S. Bryan, " Reliability Assurance - Reactor Protection System," July 23, 1981.
06/30/86 R-28 NUREG-0933 l
Ravision 3 i
449. Memorandum for S. Hanauer from D. Eisenhut, " Potential Generic Issue: BWR Control Rod Test Requirements Following Maintenance," November 26,- 1982.
450. Memorandum for R. Mattson from T. Speis, " Potential Generic Issues Related to Scram Systems," April 7, 1982.
451. Memorandum for H. Denton from C. Heltemes, " Potential Design Deficiency in Westinghouse Reactor Protection System," March 10, 1983.
452. Memorandum for C. Heltemes from H. Denton, " Westinghouse Reactor Protection System Design Conformance to IEEE Standard 279," May 2, 1983.
453. Memorandum for H. Denton et. al., from R. Mattson, " Recommended Generic i
Actions," April 27, 1983.
, 454. SECY-83-98E, " Salem Restart Evaluation," April 11, 1983.
455. NUREG-0771, " Regulatory Impact of Nuclear Reactor Accident Source Term Assumptions," U.S. Nuclear Regulatory Commission, June 1981.
456.-WASH-1248, " Environmental Survey of the Uranium Fuel Cycle," U.S. Nuclear Regulatory Commission, April 1974.
457. NUREG-0116, " Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle," U.S. Nuclear Regulatory Commission, October.1976.
I 458. NUREG-0216, "Public Comments on the Environmental Survey of the Reprocess-ing and Waste Management Portions of the LWR Fuel Cycle," U.S. Nuclear Regulatory Commission, March 1977.
459. NUREG-0252, "The Environmental Effects of Using Coal for Generating Electricity," U.S. Nuclear Regulatory Commission, June 1977.
460. NUREG/CR-1060, " Activities, Effects, and Impacts of'the Coal Fuel Cycle for a 1,000 MWe Electric Power Generating Plant," U.S. Nuclear Regulatory Commission, February 1980.
461. NUREG-0332, " Health Effects Attributable to Coal and Nuclear Fuel Cycle Alternatives," U.S. Nuclear Regulatory Commission, November 1977.
462. NUREG/CR-0022, "Need for Power: Determination in the Sta.te Decisionmaking Process," U.S. Nuclear Regulatory Commission, March 1978.
463. NUREG/CR-0250, " Regional Econometric Model for Forecasting Electricity Demand by Sector and State," U.S. Nuclear Regulatory Commission, September 1978.
464. NUREG-0555, " Environmental Standard Review Plans for the Environmental Review of Construction Permit Applications for Nuclear Power Plants,"
U.S. Nuclear Regalatory Commission, May 1979.
465. NUREG-0398, " Federal-State Cooperation in Nuclear Power Plant Licensing,"
U.S. Nuclear Regulatory Commission, March 1980.
06/30/86 R-29 NUREG-0933
Ravision 3 466. NUREG-0942, " Conducting Need-for-Power Review for Nuclear Power Plants,"
U.S. Nuclear Regulatory Commission, December 1982.
467. NUREG/CR-2423, " Mathematical Simulation of Sediment and Radionuclide Transport in Estuaries," U.S. Nuclear Regulatory Commission, November 1982.
468. NUREG/CR-2823, "A Review of the Impact of Copper Released into Marine and Estuarine Environments," U.S. Nuclear Regulatory Commission, November 1982.
469. NUREG/CR-0892, " Chronic Effects of Chlorination Byproducts on Rainbow Trout, Salmo gairdneri," U.S. Nuclear Regulatory Commission, November 1980.
470. NUREG/CR-0893, " Acute Toxicity and Bioaccummulation of Chloroform to Four Species of Freshwater Fish," U.S. Nuclear Regulatory Commission, August k 1980.
471. NUREG/CR-2750, " Socioeconomic Impacts of Nuclear Generating Stations,"
U.S. Nuclear Regulatory Commission, July 1982.
472. NUREG/CR-2861, " Image Analysis for Facility Siting: A Comparison of Low and High-Attitude Image Interpretability for Land Use/ Land Cover Mapping,"
U.S. Nuclear Regulatory Commission, November 1982.
473. NUREG/CR-2550, " Charcoal Performance Under Simulated Accident Conditions,"
U.S. Nuclear Regulatory Commission, July 1982.
474. NUREG-0700, " Guidelines for Control Room Design Reviews," U.S. Nuclear Regulatory Commission, September 1981.
475. Memorandum for W. Minners from.F. Congel, "Prioritization of Generic Issue 58, Containment Flooding," May 19, 1983.
476. Regulatory Guice 4.2, " Preparation of Environmental Reports for Nuclear Power Stations," U.S. Nuclear Regulatory Commission, July 1976.
477. NUREG/CR-2692, "An Integrated System for Forecasting Electric Energy and Load for States and Utility Service Areas," U.S. Nuclear Regulatory Commission, May 1982.
478. Regulatory Guide 1.57, " Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components," U.S. Nuclear Regulatory Commission, June 1973.
479. Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Con-tainment Following a Loss-of-Coolant Accident," U.S. Nuclear Regulatory Commission, November 1978.
480. Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," U.S. Nuclear Regulatory Commission, October 1977.
06/30/86 R-30 NUREG-0933 i
Revision 3 '
l O 481. Regulat e/ Guide 1.35, " Inservice Inspection of Ungrouted Tendons in
(] Prestressed Concrete Containment Structures," U.S. Nuclear Regulatory Commission, January 1976.
482. Regulatory Guide 1.90, " Inservice Inspection of Prestressed Concrete Containment Structures with Grouted Tendons," U.S. Nuclear Regulatory Commission, August 1977.
483. ORNL-5470, " CONCEPT-5 User's Manual," Oak Ridge National Laboratory, December 1978.
484. ORNL/TM-6467, "A Procedure for Estimating Nonfuel Operation and Maintenance Costs for Large Steam-Electric Power Plants," Oak Ridge National Labora-tory, January 1979.
485. NUREG/CR-2844, "Nonfuel Operation and Maintenance Costs for Large Steam-Electric Power Plants - 1982," U.S. Nuclear Regulatory Commission, September 1982.
486. Memorandum for Z. Rosztoczy, et. al., from W. Anderson, " Seismic Scram,"
January 20, 1983.
487. Memorandum for G. Arndt from G. Burdick, " Review of Seismic Scram Report, UCRL-53037," March 3, 1983.
488. NUREG-0610 " Draft Emergency Action Level Guidelines for Nuclear Power
/G Plants," U.S. Nuclear Regulatory Commission, September 1979.
489._NUREG/CR-2963, " Planning Guidance for Nuclear Power Plant Decontamina-tion," U.S. Nuclear Regulatory Commission, June 1983.
490. Memorandum for H. Denton from C. Michelson, " Potential Generator Missiles - Generator Rotor Retaining Rings," March 16, 1982.
1 491. NRC Letter to All Licensees of Operating Westinghouse and CE PWRs (Except Arkansas Nuclear One - Unit 2 and San Onofre Units 2 and 3), " Inadequate Core Cooling Instrumentation System (Generic Letter-No. 82-28),"
December 10, 1982.
492. Memorandum for C. Michelson from H. Denton, "H. B. Robinson RCS Leak on January 29, 1981," June 15, 1981.
493. Memorandum for C. Michelson from H. Denton, " January 19, 1981, Memorandum on Degradation of Internal Appurtenances in LWR," April 30, 1981.
494. Memorandum for C. Michelson from H. Denton, "AE0D Preliminary Report on Calvert Cliffs Unit 1 Loss of Service Water," August 5, 1981.
495. Memorandum for C. Michelson from H. Denton, " Steam Generator Overfill and Combined Primary and Secondary Blowdown," May 27, 1981.
m 496. Memorandum for H. Denton from C. Michelson, " Concerns Relating to the
\
(d Integrity of a Polymer Coating for Surfaces Inside Containment (IE Draft Bulletin No. 80-21)," August 29, 1980.
06/30/86 R-31 NUREG-0933
l l
R: vision 3 l l
l 497. Memorandum for H. Denton and V. Stello from C. Michelson, "Immediate '
Action Memo: Common Cause Failure Potential at Rancho Seco - Desiccant Contamination of Air Lines," September 15, 1981.
498. Memorandum for C. Michelson from H. Denton, "AE0D Immediate Action Memo on Contamination of Instrument Air Lines at Rancho Seco," October 26, 1981.
499. Memorandum for H. Denton, et al., from C. Michelson, " Case Study Report on San Onofre Unit 1 Loss of Salt Water Cooling Event on March 10, 1980,"
August 12, 1980.
500. Memorandum for C. Michelson from H. Denton, "NRR Comments on AEOD Final Report: Case Study Report on San Onofre Unit 1 Loss of Salt Vater Cooling Event of March 10, 1980," October 8, 1982.
501. IE Bulletin No. 79-24, " Frozen Lines," U.S. Nuclear Regulatory Commission, September 27, 1979.
502. Memorandum for H. Denton and V. Stello from C. Michelson, "Inoperability of Instrumentation Due to Extreme Cold Weather," June 15, 1981.
503. Memorandum for C. Michelson from H. Denton, "AE0D Memorandum on the Inoperability of Instrumentation Due to Extreme Cold Weather," August 14, 1981.
504. Draft Regulatory Guide and Value/ Impact Statement, Task IC 126-5,
" Instrument Sensing Lines," U.S. Nuclear Regulatory Commission, March 1982.
505. Regulatory Guide 1.151, " Instrument Sensing Lines," U.S. Nuclear Regulatory Commission, July 1983.
506. Federal Register Notice 48 FR 36029, August 8,1983.
507. Memorancun for C. Michelson from H. Denton, " Interlocks and LCO's for Redundant Class 1E Tie Breakers (Point Beach Nuclear Plant Units 1 and 2)," October 16, 1980.
508. Memorandum for F. Schroeder from L. Rubenstein, " Review of General Electric Topical Report NE00-10174, Revision 1," August 18, 1982.
509. Memorandum for C. Michelson from H. Denton, "NRR Comments on AE00 Final Report: Survey of Valve Operator-Related Events Occurring During 1978, 1979, and 1983," August 19, 1982.
510. Memorandum for C. Michelson from H. Denton, " Effects of Fire Protection System Actuation on Safety-Related Equipment," August 22, 1982.
511. "Value-Impact Analysis of Recommendations Concerning Steam Generator Tube Degradations and Rupture Events," Science Applications, Inc., February 2, 1983.
512. Memorandum for D. Eisenhut from T. Speis, " DST Prioritization of Steam Generator Requirements," May 4, 1983.
06/30/86 R-32 NUREG-0933
Revision 3
/"')
[V 513. SECY-82-186A, "Make-up Nozzle Cracking in Babcock and Wilcox (B&W)
Plants," July 23,.1982.
514. B&W Document No. 77-1140611-00, "177 Fuel Assembly Owners' Group Safe End Task Froce Report on Generic Investigation of HPI/MU Nozzle Component Cracking," Babcock and Wilcox Company, 1983.
515. Memorandum for W. Minners from D. Dilanni, " Proposed Generic Issue 'PORV and Block Valve Reliability,'" June 6, 1983.
516. Memorandem for W. Johnston and L. Rubenstein from T. Speis, " Failure of Resin Demineralizer Systems and Their Effects on Nuclear Power Plant Safety," August 6, 1982.
517. Memorandum for the Atomic Safety & Licensing Boards for: Callaway Plant, Unit 1; Comanche Peak Steam Electric Station, Units 1 & 2; and the Atomic Safety & Licensing Appeal Board for Virgil C. Summer Nuclear Station, Unit 1, from T. Novak, " Board Notification - Control Rod Drive Guide Tube Support Pin Failures at Westinghouse Plants (Board Notification
.No. 82-81)," August 16, 1982.
518. Task Interface Agreement 82-32, Rev. 1, " Detached Thermal Sleeves," U.S.
Nuclear Regulatory Commission, July 26, 1982.
519. Memorandum for W. Minners from L. Hulman, " Generic Issue on Iodine Coolant Activity Limiting Conditions for Operation," June 10, 1983.
\
520. NRC Letter to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits, " Required Actions Based on Generic Implications of Salem ATWS Events," (Generic Letter No. 83-28),
July 8, 1983.
521. SECY-83-248, " Generic Actions for Licensees and Staff in Response to the ATWS Events at Salem Unit 1," June 27, 1983.
522. AE0D/P301, " Report on the Implications of the ATWS Events at the Salem Nuclear Power Plant on the NRC Program for Collection and Analysis of Operational Experience," Office for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, July 1983.
523. Memorandum for C. Heltemes from H. Denton, "AE0D Final Report on the Implications of the ATWS Events at the Salem Nuclear Power Plant on the NRC Program for Collection and Analysis of Operational Experience,"
July 21, 1983.
524. Memorandum for R. Mattson from T. Speis, " Draft CRGR Package on A-30, DC Power," May 24, 1983.
525. Memorandum for H. Denton from C. Heltemes, " Engineering Evaluation Report, Investigation of Backflow Protection in Common Equipment and Floor Drain Systems to Prevent Flooding of Vital Equipment in Safety-O Related Compartments," March 11, 1983.
U 06/30/86 R-33 NUREG-0933
Revisisn 3 526. NUREG/CR-3226, " Station Blackout Accident Analyses (Part of NRC Task Action Plan A-44)," U.S. Nuclear Regulatory Commission, May 1983.
527. IE Information Notice No. 83-44, " Potential Damage to Redundant Safety Equipment as a Result of Backflow Through the Equipment and Floor Drain System," U.S. Nuclear Regulatcry Commission, July 1,1983.
528. Memorandum for B. Liaw from H. Berkow, "0MB Clearance Renewal - Monitoring of Fatigue Transient Limits for Reactor Coolant System," May 13, 1983.
529. Memorandum for H. Berkow from W. Minners, "0MB Clearance Renewal -
Monitoring of Fatigue Transient Limits for Reactor Coolant System,"
June 1, 1983.
530. Letter to R. DeYoung (NRC) from J. Taylor (B&W), "Unanalyzed Reactor Vessel Thermal Stress During Cooldown," March 18, 1983.
531. Memorandum for R. Vollmer from W. Minners, "B&W Notification Concerning an Unanalyzed Reactor Vessel Thermal Stress During Cooldown," April 7, 1983.
532. Regulatory Guide 1.93, " Availability of Electric Power Sources," U.S.
Nuclear Regulatory Commission, December 1974.
533. Memorandum for W. Minners from R. Bosnak, "B&W Notification Concerning an Unanalyzed Reactor Vessel Thermal Stress During Cooldown," April 26, 1983.
534. Memorandum for Chairman Palladino, et al. , from D. Eisenhut, "Unanalyzed Re, actor Vessel Thermal Stress During Cooldown (Board Notification
- BN-83-42)," April 12, 1983.
535. CE-NPSD-154, " Natural Circulation Cooldown, Task 430 Final Report,"
Combustion Engineering, Inc., October 1981.
536. B&W Document No. 86-1140819-00, " Reactor Vessel Head Cooldown During Natural Circulation Coo'.down Transients," Babcock & Wilcox Company, February 8, 1983.
537. Memorandum for W. Dircks from R. Fraley, August 18, 1982.
538. Memorandum for R. Fraley from H. Denton, "ACRS Inquiry on Pipe Break Effects on CRD Hydraulic Lines," October 29, 1982.
539. Letter to W. Dircks f-om J. Ebersole, "ACRS Comments Regarding Potential Pipe Break Effects on Control Rod Drive Hydraulic Lines in the Drywells of BWR Mark I and II Containments," March 16, 1983.
540. BNL-NUREG-28109, " Thermal-Hydraulic Effects on Center Rod Drop Accidents in a Boiling Water Reactor," Brookhaven National Laboratory, July 1980.
541. Memorandum for B. Sheron from C. Berlinger, "ACRS Request for Information Related to LOCA Effects on CRD Hydraulic Lines," October 19, 1982.
06/30/86 R-34 NUREG-0933 1
R2 vision 3 l
/O 542. Menorandum for R. Mattson, et al., from D. Eisenhut, " Potential Safety Froblems Associated With Locked Doors and Barriers in Nuclear Power Plants," May 31, 1983.
543. Memorandum for T. Speis from R. Mattson,." Proposed Generic Issue on Beyond Design Basis Accidents in Spent Fuel Pools," August 10, 1983.
544. NUREG/CR-0649, " Spent Fuel Heatup Following Loss of Water During Storage,"
U.S. Nuclear Regulatory Commission, May 1979.
545. Memorandum for Z. Rosztoczy from P. Williams, " Trip Report: International Meeting on Severe Fuel Damage and Visit to Power' Burst Facility," April 25, 1983.
546. Letter to H. VanderMolen (NRC) from D. Strenge (PNL), September 30, 1983.
547. Memorandum for W. Dircks from R. Mattson, " Closeout of NUREG-0660 Item II.E.5.1 Design Sensitivity of B&W Plants for Operating Plants 1" March 15, 1983.
548. Memorandum for W. Dircks from R. DeYoung, "TMI Action Plan Completed Items," January 26, 1983.
549. NUREG/CR-2039, " Dynamic Combinations for Mark II Containment Structures,"
U.S. Nuclear Regulatory Commission, June 1982.
550. NUREG/CR-1890, " ABS, SRSS and CDF Response Combination Evaluation for
.( Mark III Containment and Drywell Structures," U.S. Nuclear Regulatory Commission, March 1982.
551. Letter to N. Palladino from J. Ray, "Need for Rapid Depressurization Capability in Newer Combustion Engineering,.Inc. Plants," October 18, 1983.
552. Memorandum for W. Minners from B. Siegel, " Proposed Generic Issue
' Reliability of Vacuum Breakers Connected to Steam Discharge Lines Inside BWR Containments,'" November 3, 1983.
553. Memorandum for D. Eisenhut from J. 01shinski, " Loss of High Head Injection Capability at McGuire Unit 1 and Reconsideration of Technical Specifica-tion 3.0.3 and 3.5.2," April 12, 1982.
554. Memorandum for D. Eishenhut, et al., from H. Denton, " Development of Generic Recommendations Based on the Review of the January 25, 1982 Steam Generator Tube Rupture at Ginna," May 3, 1982.
555. Letter to D. Eishenhut from D. Waters (BWR Owners' Group), "BWR Owners' Group Evaluations of NUREG-0737 Requirements II.K.3.16 and II.K.3.18,"
March 31, 1981.
556. Memorandum for G. Lainas, et al. , from W. Houston, " Evaluation of BWR Owners' Group Generic Response to Item II.K.3.16 of NUREG-0737, ' Reduction p)
(u of Challenges and Failures of Relief Valves - Feasibility Study and System Modification,'" April 1, 1983.
06/30/86 R-35 NUREG-0933
Revision 3 557. Memorandum for H. Denton and V. Stello from C. Michelson, "Calvert Cliffs Unit 1 Loss of Service Water," June 19, 1981.
558. Memorandom for H. Denton and R. DeYoung from C. Michelson, "Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980," December 17, 1981.
559. Memorandum for C. Michelson from H. Denton, "NRR Comments on AE0D Final Report: Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980,"
September 23, 1982.
560. Memorandum for H. Denton from C. Heltemes, " Response to NRR Comments on-AE0D Report, 'Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980,'"
May 2, 1983.
561. Memorandum for W. Houston and L. Rubenstein from F. Miraglia, " Response to NRR Comments on AE0D Report, 'Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980,'" June 2, 1983.
562. Memorandum for F. Miraglia from W. Houston and L. Rubenstein, " Comments to AE00 Memo dated May 2, 1983, on Calvert Cliffs, Unit 1, Loss of Service Water on May 20,1980," July 22, 1983.
563. Memorandum for C. Heltemes from H. Denton, " Response to NRR Comments on AE0D Report, 'Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980,'"
September 15, 1983.
564. Memorandum for R. Baer from K. Seyfrit, " Case Study, 'Calvert Cliffs Unit 1 Loss of Service Water on May 29, 1980,'" August 18, 1983.
565. IE Information Notice No. 83-77, " Air / Gas Entrainment Events Resulting in System Failures," U.S. Nuclear Regulatory Commission, November 14, 1983.
566. Memorandum for G. Holahan from W. Minners, "Prioritization of Issue 36:
Loss of Service Water at Calvert Cliffs Unit 1," November 10, 1983.
567. Letter to A.E. Lundvall (Baltimore Gas and Electric Company) from D. Eisenhut (NRC), Docket No. 50-317, September 15, 1983.
568. Memorandum for W. Houston and L. Rubenstein from F. Schroeder, " Request for Reactor Systems Branch and Auxiliary Systems Branch Support'for Plant Visits on USI A-45," November 28. 1983.
569. " Engineering Evaluation of the H. B. Robinson Reactor Coolant System Leak on January 29, 1981," Office for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, March 23, 1981.
570. Memorandum for V. Stello from H. Denton, " Issuance of Revised Section 7.1, Appendix A to this Section, Section 7.5 and Section 7.7 of the Standard Review Plan, NUREG-0800," March 9, 1984.
571. Memorandum for H. Denton from V. Stello, "SRP Changes Concerning Resolt. tion of Generic Issue 45, Inoperability of Instrumentation due to Extreme Cold Weather," April 3, 1984.
06/30/86 R-36 NUREG-0933
Revision 3 h-
/g (V ) 572. Memorandum for G. Lainas from F. Rowsome, " Safety Evaluation of the Westinghouse Licensees' Responses to TMI Action Item II.K,3.2," July 22, 1983.
573. Memorandum for G. Lainas from F. Rowsome, " Safety Evaluation of the B&W Licensees' Responses to TMI Action Item II.K.3.2," August 24, 1983.
574. Memorandum for G. Lainas from F. Rowsome, " Safety Evaluation of the CE Licensees' Responses to TMI Action Item II.K.3.2," August 26, 1983.
575. Memorandum for W. Minners from'B. Sheron, " Proposed Generic Issue on PORV and Block Valve Reliability," June 27, 1983.
576. Memorandum for R. Riggs from F. Cherny, " Comments on Draft Write-up of Prioritization of Generic Issue 70 'PORV and Block Valve Reliability,'"
December 21, 1983.
577. Memorandum for H. Denton, et al., from C. Heltemes, " Case Study Report -
Low Temperature -0verpressure Events at Turkey Point Unit 4," September 26, 1983.
578. NUREG-0748, " Operating Reactors Licensing Actions Summary," U.S. Nuclear Regulatory Commission.
579. NUREG-0694, "TMI-Related Requirements for New Operating Licenses," U.S.
Nuclear Regulatory Commission, June 1980.
\v/ 580. NUREG-0645, " Report of the Bulletins and Orders Task Force," U.S. Nuclear Regulatory Commission, January 1980.
581. NUREG-0909, "NRC Report on the January 25, 1982 Steam Generator Tube Rupture at R.E. Ginna Nuclear Power Plant," U.S Nuclear Regulatory Commis-sion, April 1982.
582. NUREG-0713, " Occupational Radiation Exposure at Commerical Nuclear Power Reactors-1981," U.S Nuclear Regulatory Commission, November 1982.
583. EPRI NP-2292, "PWR Safety and Relief Valve Test Program," Electric Power Research Institute, December 1982.
584. EPRI NP-1139, " Limiting Factor Analysis of High Availability Nuclear Plants, Electric Power Research Institute, August 1979.
585. EPRI P-2410-SR, " Technical Assessment Guide," Electric Power Research Institute, May 1982.
586. WCAP-9804, "Probabilistic Analysis and Operational Data in Response to NUREG-0737, Item II.K.3.2 for Westinghouse NSSS Plants," Westinghouse Electric Corporation, February 1981.
587. " Accident Sequence Evaluation Program, Phase II Workshop Report," Sandia i National Laboratories, EG&G Idaho, Inc., and Science Applications, Inc.,
Q September 1982.
06/30/86 R-37 NUREG-0933
l l
1 Revision 3 I 1
l J
588. Letter to Director, NRR from K. Cook (Louisiana Power & Light), "Waterford SES Unit 3, Docket No. 50-382, Depressurization and Decay Heat Removal,"
October 27, 1983.
589. Letter to W. Dircks from E. Van Brunt, Jr. (Arizona Public Serivce Company),
"Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3, Docket Nos. STN-50-528/529/530," November 7, 1983.
590. ALO-75 (TR-3459-1), " Pilot Program to Identify Valve Failures Which Impact the Safety and Operation of Light Water Nuclear Power Plants," Teledyne Engineering Services, January 11, 1980.
591. IE Information Notice No. 82-45, "PWR Low Temperature Overpressure Protection," U.S. Nuclear Regulatory Commission, November 19, 1982.
592. IE Information Notice No. 82-17, "0verpressurization of Reactor Coolant System," U.S. Nuclear Regulatory Commission, June 10, 1982.
593. SECY-84-76, " Proposed Rulemaking for Operator Licensing and for Training and Qualifications of Civilian Nuclear Power Plant Personnel," February 13, 1984.
594. Letter to E. Wilkinson (INPO) from W. Dircks (NRC), November 23, 1983.
595. SECY-83-52A, " Final Rulemaking Concerning Licensed Operator Staffing at Nuclear Power Units and Draft Policy Statement on Shift Crew Qualifications,"
March 14, 1983.
596. Memorandum for W. Dircks from S. Chilk, " Staff Requirements - Affirmation /
Discussion and Vote, 3:35 p.m., Thursday, April 21, 1983, Commissioners' Conference Room (0 pen to Public Attendance)," April 28, 1983.
l 597. Federal Register, Vol. 48, No. 144, pp. 33850-33860, " Licenser Event Report System," July 26, 1983.
598. Memorandum for W. Dircks from H. Denton, " Closeout of TMI Action Plan Task III.D.2.5, 'Offsite Dose Calculation Manual,'" January 17, 1984.
599. NUREG/CR-3332, " Radiological Assessment - A Textbook on Environmental Dose Analysis," U.S. Nuclear Regulatory Commission, September 1983.
600. NUREG-0978, " Mark III LOCA-Related Hydrodynamic Load Definition," U.S.
Nuclear Regulatory Commission, February 1984.
601. Memorandum for T. Combs from H. Denton, " Revised SRP Section 6.2.1.1.C of NUREG-0800," September 10, 1984.
602. Memorandum for T. Speis from R. Mattson, " Status of Generic Issues 40 and 65 Assigned to DSI," December 27, 1983.
603. NUREG-0985, "U.S. Nuclear Regulatory Commission Human Factors Program Plan,"
U.S. Nuclear Regulatory Commission, August 1983.
R-38 NUREG-0933 06/30/86
1 Ravision 3 1 l
604. SECY-81-641, " Electromagnetic Pulse (EMP) - Effects on Nuclear Power b Plants," November 5, 1981.
~605. SECY-82-157, " Status Report on the Evaluation of the Effects of Electro-magnetic Pulse (EMP) on Nuclear Power Plants," April 13, 1982.
606.'SECY-83-367, " Staff Study of Electromagnetic Pulse (EMP) Effects on Nuclear Power Plants and Discussion of Related Petitions for Rulemaking (PRM-50-32, 32A, and 32B)," September 6, 1983.
607. Memorandum for W. Dircks from S. Chilk, "SECY-83-367 - Staff Study of Electromagnetic Pulse (EMP) Effects on Nuclear Power Plants and Discussion of Related Petition for Rulemaking (PRM-50-32, 32A, and 32B)," November 15, 1983.
608. IE Information Notice No. 82-39, " Service Degradation of Thick-Walled Stainless Steel Recirculation Systems at BWR Plants," U.S. Nuclear Regula-
. tory Commission, September 21, 1982.
609. IE Bulletin No. 82-03, " Stress Corrcsion Cracking in Thick-Wall Large Diameter, Stainless Steel Recirculation System Piping at BWR Plants,"
U.S. Nuclear Regulatory Commission, October 14, 1982.
610. IE Bulletin No. 83-02, " Stress Corrosion Cracking in large Diameter Stain-less Steel Recirculation System Piping at BWR Plants," U.S. Nuclear Regula-tory Commission, March 4, 1983.
611. NUREG-1061, " Report of the U.S. Nuclear Regulatory Commission Piping Review Committee," U.S. Nuclear Regulatory Commission, (Vol. 1) August 1984, (Vol. 2) April 1985, (Vol. 3) November 1984, (Vol. 4) December 1984, (Vol. 5) April 1985.
612. SECY-83-267, " Status Report on Observation of Pipe Cracking at BWRs,"
July 1, 1983.
613. SECY-83-267A, " Update of Status Report on Observation of Pipe Cracking at BWRs (SECY-83-267)," July 11, 1983. +
614. SECY-83-267B, " Update of Status Report on Observation of Pipe Cracking at BWRs (SECY-83-267 and 267A)," August 8, 1983.
4 615. SECY-83-267C, " Staff Requirements for Reinsp'ection of BWR Piping and Repair of Cracked Piping," November 7, 1983.
616. SECY-84-9, " Report on the Long Term Approach for Dealing with Stress Corrosion Cracking.in BWR Piping," January 10, 1984.
617. SECY-84-9A, " Update of Status Report on BWR Pipe Cracks and Projection of Upcoming Licensee Actions," January 27, 1984.
618. SECY-84-166, " Update of Status Report on BWR Pipe Cracks and Projection of p Upcoming Licensee Actions," April 20, 1984.
U 619. SECY-84-301, " Staff Long Range Plan for Dealing with Stress Corrosion Cracking in BWR Piping," July 30, 1984.
06/30/86 R-39 NUREG-0933
Revision 3 620. NRC Letter to All Licensees of Operating Reactors, Applicants for Operating License, and Holders of Construction Permits for Boiling Water Reactors,
" Inspections of BWR Stainless Steel Piping," (Generic Letter 84-11),
April 19, 1984.
621. " Report to the NRC Office of Nuclear Material Safety and Safeguards,"
Committee to Review Safeguards Requirements at Power Reactors, U.S.
Nuclear Regulatory Commission, February 28, 1983.
622. Memorandum for T. Speis from R. Mattson, " Fuel Crubling During LOCA,"
February 2, 1983.
623. Memorandum for H. Denton from D. Eisenhut, " Potential Safety Problems Associated with Locked Doors and Barriers in Nuclear Poster Plants,"
December 22, 1983.
624. Memorandum for D. Eisenhut from H. Denton, " Safety-Safeguards Interface,"
January 16, 1984.
625. Memorandum for H. Thompson from D. Eisenhut, " Potential Safety Problems Associated with Locked Doors and Barriers in Nuclear Power Plants,"
January 30, 1984.
626. Memorandum for T. Speis from H. Thompson, " Submittal of Potential Gerieric Issue Associated with Locked Doors and Barriers," June 8, 1984.
627. SECY-83-311, " Proposed Insider Safeguards Rules," July 29, 1983.
628. IE Information Notice No. 83-36, " Impact of Security Practices on Safe Operations," U.S. Nuclear Regulatory Commission, June 971983.
629. Memorandum for the Record from L. Bush, " Probability of Failure of Locks,"
May 24, 1984.
630. Memorandum for W. Minners from F. Miraglia, " Proposed Generic Issue -
Technical Specifications for Anticipatory Trips," February 23, 1984.
631. Memorandum for F. Miraglia from W. Houston, " Task Interface Agreement Task No. 83-77 (TAC 40002, PA-157)," November 29, 1983.
632. Memorandum for D. Ross from H. Richings, "RDA Statistical Analysis,"
June 17, 1975.
633. Memorandum for P. Check from H. Richings, "Some Notes on PWR (W) Power Distribution Probabilities for LOCA Probabilistic Analyses," July 5,1977.
634. NUREG-0630, " Cladding Swelling and Rupture Models for LOCA Analysis,"
U.S. Nuclear Regulatory Commission, April 1980.
635. Memorandum for G. Holaham and W. Minners from R. Mattson, " Disposition of AE00 Engineering and Technical Evaluation Reports," April 10, 1984.
636. Memorandum for R. DeYoung and H. Denton from C. Heltemes, " Vapor Binding of Auxiliary Feedwater Pumps," November 21, 1983.
06/30/86 R-40 NUREG-0933
Revision 3
[m V] 637. AE0D/C404, " Steam Binding of Auxiliary Feedwater Pumps," Office for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, July 1984.
638. Memorandum for H. De'nton from C. Michelson, " Tie 8reaker Between Redundant Class 1E Buses - Point Beach Nuclear Plant, Units 1 and 2," August 27, 1980.
639. Letter to J. Keppler (NRC) from C. Fay (Wisconsin Electric Power Company),
" Docket No. 50-301, Point Beach Nuclear Plant Unit 2 Licensee Event Report ~
No. 80-005/03L-0," June 27, 1980.
640. Memorandum for H. Denton from C. Heltemes, "Special Study Report - Human Error in Events Involving Wrong Unit or Wrong Train," January 13, 1984.
641. IE Information Notice No. 84-51, " Independent Verification," U.S. Nuclear Regulatory Commission, June 26, 1984.
642. IE Information Notice No. 84-58, " Inadvertent Defeat of Safety Function Caused by Human Error Involving Wrong Unit, Wrong Train, or Wrong System,"
j U.S. Nuclear Regulatory Commission, July 25, 1984~.
643. Memorandum for H. Denton from C. Heltemes, " Human Error in Events Involving Wrong Unit or Wrong Train," August 8, 1984.
n 644. Memorandum for D. Eisenhut, et al. , from H. Thompson, " Maintenance and Surveillance Program Implementation Plan," July 7,1984.
645. Memorandum for C. Heltemes from H. Denton, "Special Study Report - Human Errors in Events Involving Wrong Unit or Wrong Train," May 2, 1984.
646. Memorandum for C. Heltemes from H. Denton, " Human Error in Events Involving Wrong Unit or Wrong Train," September 17, 1984.
647. Memorandum for T. Speis from H. Denton, " Resolution of Generic Issue B-26,
' Structural Integrity of Containment Penetrations,'" September 27, 1984.
648. Memorandurr for T. Speis from H. Dentcn, "Closecut of Generic Issue B-54,
' Ice Condenser Containments,'" October 22, 1984.
649. NUREG/CR-3716, " CONTEMPT 4/ MOD 4," U.S. Nuclear Regulatory Commission, March 1984.
650. NUREG/CR-4001, " CONTEMPT 4/ MOD 5," U.S. Nuclear Regulatory Commission, September 1984.
651. NUREG-0985, Revision 1, "U.S. Nuclear Regulatory Commission Human Factors Program Plan," U.S. Nuclear Regulatory Commission, September 1984.
652. Memorandum for W. Dircks from R. DeYoung, " Elimination of Duplicative Tracking Requirements for Revision of Regulatory Guide 1.33," July 26, 1984.
O 06/30/86 R-41 NUREG-0933
Revisien 3 653. NUREG/CR-3123, " Criteria for Safety-Related Nuclear Power Plant Operator Actions: 1982 Pressurized Water Reactor (PWR) Simulator Exercises," U.S.
Nuclear Regulatory Commission, June 1983.
654. Memorandum for W. Dircks from H. Thompson, " Closeout of TMI Action Plan Task I.G.2, ' Scope of Test Program,'" October 5, 1984.
655. Memorandum for W. Dircks from H. Denton, " Generic Issue II.A.1, ' Siting Policy Reformulation,'" September 17, 1984.
656. Memorandum for W. Dircks from H. Denton, " Closeout of TMI Action Plan Task II.E.5.2, Transient Response of B&W Designed Reactors," September 23, 1984.
657. Memorandum for D. Crutchfield from D. Eisenhut, "TMI Action Plan Task II.E.5.2," November 6, 1984.
658. NUREG-1054, " Simplified Analysis for Liquid Pathway Studies," U.S. Nuclear Regulatory Commission, August 1984.
659. Memorandum for H. Denton from R. Vollmer, "ESRP 7.1.1 ' Environmental -
Impacts of Postulated Accidents Involving Radioactive Materials - Releases to Groundwater,'" September 25, 1984.
660. Memorandum for W. Dircks from H. Denton, " Generic Issue III.D.2.3 ' Liquid Pathway Radiological Control,'" October 29, 1984.
661. Memorandum for H. Denton from C. Heltemes, " Failures of Class 1E Safety-Related Switchgear Circuit Breakers to Close on Demand," April 29, 1983.
652. Memorardum for C. Heltemes from H. Denton, "AE0D April 1983 Report on Failures of Class 1E Safety-Related Switch Gear Circuit Breakers to Close on Demand," June 17, 1983.
663. IE Information Notice No. 83-50, " Failures of Class 1E Safety-Related Switchgear Circuit Breakers to Close on Demand," August 1, 1983.
664. Memorandum for D. Eisenhut from R. Spessard, "Unmonitored Failures of Class 1E Safety-Related Switchgear Circuit Breakers and Power Supplies (AITS-F03052383)," June 1, 1984.
665. NUREG/CR-2989, " Reliability of Emergency AC Power System at Nuclear Power Plants," U.S. Nuclear Regulatory Commission, July 1983.
666. Memorandum for T. Speis from H. Denton, " Resolution of Generic Issue B-12:
BWR Jet Pump Integrity," September 25, 1984.
667. Memorandum for T. Speis from H. Denton, " Resolution of Generic Issue 69:
Make-up Nozzle Cracking in B&W Plants," September 27, 1984.
668. Memorandum for H. Denton from R. Minogue, " Comments on Generic Issue 79,
'Unanalyzed Reactor Vessel Thermal Stress During Natural Convection Cooldown,'" October 5, 1983.
06/30/86 R-42 NUREG-0933
,i.
I
{ Revision 3 i
! [m 669. Letter to P. Kadambi (NRC) from F. Miller (B&W Owners Group Analysis
! Committee), " Transmittal of RV Head Stress Evaluation Program Results,"
October 15, 1984.
I
- 670. Memorandum for H. Denton from R. Mattson, " Generic Issue B-60, Loose Parts
] Monitoring Systems for Operating Reactors (TACS 52325)," January 10, 1984.
i j 671. Letter to N. Palladino from P. Shewmon, " Control Room Habitability," ;
August 18, 1982. ;
i
- 672. Letter to J. Ray from W. Dircks, " August 18, 1982 ACRS Letter on Control
{ Room Habitability," January 31, 1983.
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.
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 j Habitability," September 19, 1983.
677. Letter to W. Milstead (NRC) from T. Powers (PNL), December 3, 1984.
\
678. Memorandum for W. Dircks from H. Denten, " Control Room Habitability,"
June 29, 1984.
679. Memorandum for T. Speis from R. Bernero, " Revised Schedule for Generic i
Issue 83, Control Room Habitability," September 28, 1984.
4 I 680. NUREG/CR-2258, " Fire Risk Analysis for Nuclear Power Plants," U.S.
j Nuclear Regulatory Commission, September 1981.
- 681. NUREG-0844, "NRC Integrated Program for the Resolution of Unresolved Safety l Issues A-3, A-4, A-5 Regarding Steam Generator Tube Integrity," U.S. Nuclear Regulatory Comm;ssion, (Draf t) December 1983, (Draf t) April 1985.
- 682. Note to W. Kane from G. Holahan, " Background Information Relating to the i . Assessment of the Offsite Consequences of Non-Core Melt, Steam Generator 2
Tube Rupture Events," October 24, 1983.
683. Memorandum for W. Johnston from R. Ballard, " Disputed Procedures for Estimating Probable Maximum Precipitation," January 13, 1984.
l' 684. Hydrometeorological Report No. 52, " Application.of Probable Maximum Precipitation Estimates - United States East of the 105th Meridian," U.S.
! Department of Commerce, National Oceanic and Atmospheric Administration, i August 1982.
.y .
685. Hydrometeorological Report No. 51, " Probable Maximum Precipitation
- Estimates, United States East of the 105th Meridian," U.S. Department of Commerce, National Oceanic and Atmospheric Administration, June 1978.
F j .06/30/86 R-43 NUREG-0933 t
, - , - , - . - ,r -,.. , - - ,. , ,y--- r..y.-- - - .%---m,+,r-rm,w-,m,-we
- -c-,,-,-.. - ---.--. . , . . , _ _ _,--,,,-w.m , , ,www.mw,,,, , , , , - .- ~ . - -
Revision 3 686. Hydrometeorological Report No. 33, " Seasonal Variation of the Probable Maximum Precipitation East of the 105th Meridian for Areas from 10 to 1,000 Square Miles and Durations of 6, 12, 24 and 48 Hours," U.S.
Department of Commerce, April 1956.
687. Regulatory Guide 1.59, " Design Basis Floods for Nuclear Power Plants,"
U.S. Nuclear Regulatory Commission, August 1977.
688. Regulatory Guide 1.102, " Flood Protection for Nuclear Power Plants," U.S.
Nuclear Regulatory Commission, SepterrSer 1976.
689. Memorandum for V. Stello from H. Denton, " Potential Generic Requirement Concerning Design for Probable Maximum Precipitation," June 25, 1984.
690. Memorandum for V. Stello from H. Denton, " Generic Requirements Regarding Design for Probable Maximum Precipitation," October 10, 1984.
691. Memorandum for H. Denton from V. Stello, " Generic Requirements Regarding Design for Probable Maximum Precipitation," August 8, 1984.
692. Memorandum for T. Speis from H. Denton, " Generic Issue A-41; ' Long Term Seismic Program,'" October 10, 1984.
693. Memorandum for H. Denton from R. Bernero, " Resolution of Generic Issue No. 22, Inadvertent Baron Dilution Events (BDES)," September 17, 1984.
694. Memorandum for T. Speis from H. Denton, " Closeout of Generic Issue No. 22,
' Inadvertent Boron Dilution Events (BDE),'" October 15, 1984.
695. Memorandum for T. Speis from H. Denton, " Closeout of Generic Issue 50,
' Reactor Vessel Level Instrumentation in BWRs,'" October 17, 1984.
696. NRC. Letter to All Boiling Water Reactor (BWR) Licensees of Operating Reactors (Except Lacrosse, Big Rock Point, Humboldt Bay and Dresden-1),
" Reactor Vessel Water Level Instrumentation in BWRs (Generic Letter No. 84-23," October 26, 1984.
697. Memorandum for D. Eisenhut from R. Bernero, " Resolution of Generic Issue 50, Reactor Vessel Level Instrumentation O. BWRs," September 6, 1984.
698. NUREG-0927, Revision 1, " Evaluation of Water Hammer Occurrence in Nuclear Power Plants," U.S. Nuclear Regulatory Commission, March 1984.
699. NUREG-0609, " Asymmetric Blowdown Loads on PWR Primary Systems," U.S.
Nuclear Regulatory Commission, January 1981.
700. NRC Letter to All Operating PWR Licenses, Construction Permit Holders, and Applicants for Construction Permits, " Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops (Generic Letter 84-04)," February 1, 1984.
701. NUREG-0408, " Mark I Containment Short-Term Program Safety Evaluation Report," U.S. Nuclear Regulatory Commission, December 1977.
06/30/86 R-44 NUREG-0933
/
- Revision 3
[ 702. NUREG-0661, " Mark I Coritainment Long Term Program Safety Evaluation Q Report, Resolution of Generic Technical Activity A-7," U.S. Nuclear Regulatory Commission, July 1980.
NUREG-0661, Supplement 1, " Safety Evaluation Report for the Mark I Containment Long-Term Program," U.S. Nuclear Regulatory Commission, August 1982.
703. NUREG-0808, " Mark II Containment Program Evaluation and Acceptance Criteria," U.S. Nuclear Regulatory Commission, August 1981.
704. NUREG-0460, " Anticipated Transients Without Scram for Light Water Reactors," U.S. Nuclear Regulatory Commission, March 1980.
705. Memorandum for C. Thomas from O. Parr, "CRD Accumulators - Proposed Improved Technical Specification," August 13, 1984.
706. NUREG-0123, Revision 3, " Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)," U.S. Nuclear Regulatory Commis-sion, December 1980.
707. Memorandum for H. Denton, et. al. , from C. Michelson, " Survey of Valve Operator-Related Events Occurring During 1978, 1979, and 1980," December 23, 1981.
708. Memorandum for C. Michelson from H. Denton, "NRR Comments on AEOD Draft i Report: . Survey of Valve Operator-Related Events Occurring During 1978, (d 1979 and 1980," May 5, 1982.
709. AE0D/C203, " Survey of Valve Operator-Related Events Occurring During 1978, 1979, and 1980," Office- for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, May 1982.
710. Memorandum for C. Michelson from E. Brown and F. Ashe, "AE0D Assessment of Program Office Responses to the Report AE0D/C203, ' Survey of Valve Operator-Related Events Occurring During 1978, 1979, and 1980,'"
December 23, 1982.
711. Memorandum for H. Denton from C. Michelson, "AE0D Assessment of Program Office Responsea to AE00 Case Study (C-203), ' Survey of Valve Operator Related Events Occurring During 1978, 1979, and 1980,'" January 12, 1983.
712. Memorandum for C. Michelson from H. Denton, "AE0D Assessment of Program Office Responses to AEOD Case Study (C203), ' Survey of Valve Operator Related Events Occurring Durir.g 1978, 1979, and 1980,'" February 23, 1983.
713. Memorandum for K. Seyfrit from E. Brown and F. Ashe, " Engineering Evalua-tion Report AE0D/E305 Inoperable Motor Operated Valve Assemblies Due to Premature Degradation of Motors and/or Improper Limit Switch / Torque Switch Adjustment," April 13, 1983.
714. Memorandum for W. Minners from R. Bosnak, " Status of Potential Generic Issue 54, ' Valve Operator Related Events Occurring During 1978, 1979, and 1980,'" March 26, 1984.
06/30/86 R-45 NUREG-0933
R; vision 3 715. Memorandum for R. Vollmer from R. Bosnak, "MEB Task Action Plan for Resolution of Generic Issue II.E.6.1, 'In Situ Testing of Valves,'"
July 30, 1984.
716. Memorandum for D. Eisenhut from D. Muller, "PWR Reactor Cavity Uncontrolled Exposures, Generic Letter Implementina a Generic Technical Specification,"
July 12, 1984.
717. Memorandum for A. Thadani from W. Minners, "CRAC2 Computer Runs in Support of USI A-43," February 1, 1983.
718. Memorandum for W. Minners from F. Congel, "Prioritization of Generic Issue 97: PWR Reactor Cavity Uncontrolled Exposures," February 8,1985.
719. Memorandum for H. Denton from R. Bernero, "PWR Reactor Cavity Uncontrolled Exposures," November 28, 1984.
720. Memorandum for T. Speis from R. Bernero, " Request for Prioritization of Generic Safety Issue - Break Plus Single Failure in BWR Water Level Instru-mentation," October 10, 1984.
721. Memorandum for H. Denton and V. Stello from C. Michelson, " Case Study Report - Safety Concern Associated with Reactor Vessel Instrumentation in Boiling Water Reactors," September 2, 1981.
722. Memorandum for B. Sheron from A. Thadani, " Reactor Vessel Level Instru-mentation in BWR's (Generic. Issue 50)," August 2, 1984.
{
723. Memorandum for H. Denton from T. Speis, " Reactor Vessel Level Instrumenta-tion in BWRs (Generic Issue 50)," August 2, 1984.
724. Memorandum for W. Dircks, et al. , from S. Chilk, " Staff Requirements -
Af firmation/ Discussion and Vote,11:30 a.m. , Friday, June 1,1984, Com-missi(ners' Conference Room, D.C. Office (0 pen to Public Attendance),"
June 1, 1984.
725. Federal Register, Vol. 49, No. 124, pp. 26036-26045, "10 CFR Part 50, Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants," June 26, 1984.
726. NE00-21506, " Stability and Dynamic Performance of the General Electric Boiling Water Reactor," General Electric Company, January 1977.
727. Memorandum for D. Crutchfield from L. Rubenstein, " Staff Evaluation of GE Topical Report NEDE-24011 (GESTAR) Amendment 8," April 17, 1985.
728. XN-NF-691(P)(A) & Supplement 1, " Stability Evaluation of Boiling Water Reactor Cores Sensitivity Analyses & Benchmark Analysis," Exxon Nuclear Company, Inc., August 22, 1984.
729. Memorandum for D. Eisenhut from R. Mattson, " Board Notification - BWR Core Thermal Hydraulic Stability," February 27, 1984.
06/30/86 R-46 NUREG-0933 m
Revision 3
[,~,} - 730.. Memorandum for T. Novak from L. Rubenstein, "Susquehanna 1 and 2 - Thermal v Hydraulic Stability Technical Specification Change (TACS 55021 and 55022),"
July 11, 1984.
731. Memorandum for G. Lainas'from L. Rubenstein, "SER Input for Peach Bottom-3 Technical Specification Changes for Cycle 6 Operation with Increased Core Flows and Decreased Feedwater Temperatures (TACS #55123)," October 23, 1984.
732. NEDO-21078, " Test Results Employed by GE for BWR Containment and Vertical Vent Loads," General Electric Company, October 1975.
733. NUREG-0487, " Mark II Lead Plant Program Evaluation Report," U.S. Nuclear Regulatory Commission, November 1978.
NUREG-0487, Supplement 1, " Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria," U.S. Nuclear Regulatory Commission, September 1980.
734. NUREG-0783, " Suppression Pool Temperature Limits for BWR Containments,"
U.S. Nuclear Regulatory Commission, November 1981.
735. Letter to T. Novak (NRC) from T. Pickens (BWR Owners' Group), " Agreements from BWROG/NRC Meeting on Suppression Pool Temperature Limit," October 16, 1984.
in 736. Memorandum for T. Speis from R. Bernero, " Proposed Generic Issue 'BWR
' (j\
\ Suppression Pool Temperature Limits,'" November 21, 1984.
737. Memorandum for W. Minners from W. Butler, " Comments on Prioritization of Generic Issue 108, 'BWR Suppression Pool Temperature Limits,'" January 10, 1985.
738. NUREG-1044, " Evaluation of the Need for a Rapid Depressurization Capability for CE Plant," U.S. Nuclear Regulatory Commission, December 1984.
739. SECY-84-134, " Power Operated Relief Valves for Combustion Engineering Plants," March 23, 1984.
740. " Draft Maintenance Program Plan," U.S. Nuclear Regulatory Commission, May 8, 1984.
741. NUREG/CR-3543,'" Survey of Operating Experience from LERs to Identify Aging Trend," U.S. Nuclear Regulatory Commission, January 1984.
742. NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," U.S. Nuclear Regulatory-Commission, November 1980.
743. NUREG-0744, Revision 1, " Resolution of the Task A-11 Reactor Vessel Materials Toughness Safety Issue," U.S. Nuclear Regulatory Commission, October 1982.
744. NRC Letter to All Power Reactor Licensees (Except Ft. St. Vrain),
"NUREG-0744 Rev. 1; Generic Letter No. 82-26) - Pressure Vessel Material Fracture Toughness," November 12, 1982.
06/30/86 R-47 NUREG-0933 1
+, - - - - . -- --- - -
Revision 3 745. NUREG-0885, Issue 3, "U.S. Nuclear Regulatory Commission Policy and Planning Guidance," U.S. Nuclear Regulatory Commission, January 1984.
746. NUREG-0224, " Final Report on Reactor Vessel Pressure Transient Protection for Pressurized Water Reactors," U.S. Nuclear Regulatory Commission, September 1978.
747. NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants Resolution of Generic Technical Activity A-36, "U.S. Nuclear Regulatory Commission, July 1980.
748. NUREG-0763, " Guidelines for Confirmatory Inplant Tests of Safety Relief Valve Discharges for BWR Plants," U.S. Nuclear Regulatory Commission, May 1981.
749. NUREG-0802, " Safety / Relief Valve Quencher Loads: Evaluation for BWR Mark II and III Containments," U.S. Nuclear Regulatory Commission, October 1982.
750. NUREG-0313, " Technical Report on Material Selection and Processing Guide-lines for BWR Coolant Pressure Boundary Piping," U.S. Nuclear Regulatory Commission, (Rev. 1) July 1980.
751. WASH-1270, " Anticipated Transients Without Scram for Water-Cooled Reactors,"
U.S. Nuclear Regulatory Commission, September 1973.
752. Memorandum for S. Hanauer from D. Eisenhut, "Value/ Impact Assessment of Proposed Steam Generator Generic Requirements," October 12, 1982.
753. SECY-84-13, "NRC Integrated Program for the Resolution of Steam Generator USI's," January 11, 1984.
754. NUREG-0916, " Safety Evaluation Report Related to Restart of R.E. Ginna Nuclear Power Plant," U.S. Nuclear Regulatory Commission, May 1982.
755. NUREG-0651, " Evaluation of Steam Generator Tube Rupture Events," U.S.
Nuclear Regulatory Commission, March 1980.
756. Memorandum for D. Eisenhut from T. Speis, "Prioritization of Staff Actions Concerning S.G. Tube Degradation and Rupture Events," February 23, 1983.
757. SECY-84-13A, "NRC Integrated Program for the Resolution of Steam Generator USIs," September 7, 1984.
758. SECY-84-138, "NRC Integrated Program for the Resolution of. Steam Generator USI's - Response to Commissioner Comments (Memo from Chilk to Dircks dated September 13, 1984)," November 5, 1984.
759. AE0D/C005, "AE00 Observations and Recommendations Concerning the Problem of Steam Generator Overfill and Combined Primary and Secondary Blowdown,"
Office for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, December 17, 1980.
06/30/86 R-48 NUREG-0933
Rzvision 3
[m(j} 760. NUREG/CR-2802 and Appendices A, B, and C, " Interim Reliability Evaluation Program: Analysis of the Browns Ferry, Unit 1, Nuclear Plant," U.S.
Nuclear Regulatory Commission, August 1982.
761. AE0D/E414, " Stuck Open Isolation Check Valve on the Residual Heat Removal System at Hatch Unit 2," Office for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, May 31, 1984.
762. Memorandum for W. Minners from G. Holahan, "Prioritization of Interfacing System LOCA at Boiling Water Reactors," October 25, 1984.
763. NUREG-0677, "The Probability of Intersystem LOCA: Impact Due to Leak
, Testing and Operational Changes," U.S. Nuclear Regulatory Commission, May 1980.
764. SECY-85-129, " Maintenance and Surveillance Program Plan," April 12, 1985.
765. SECY-85-62, "NRC Integrated Program for the Resolution of Steam Generator USI's - Response to Commissioner Comments (Memo from Chilk to Dircks Dated January 23, 1985)," February 22, 1985.
766. Memorandum for W. Dircks from S. Chilk, "SECY-85 NRC Integrated Pro-gram for the Resolution ~of Steam Generator USIs - Response to Commissioners Comments (Memo from Chilk and Dircks Dated January 23, 1985)," March 15, 1985.
767. Memorandum for W. Dircks from H. Denton, " Final Rule - Applicability of
(.) License Conditions and Technical Specifications in an Emergency,"
February 17, 1983.
768. Memorandum for T. Speis. from H. Denton, " Formation of a Technical Specifi-cation Improvement Project Group," December 31, 1984.
769. Memorandum for V. Stello from H. Denton, "Close Out Generic Issue #B Thermal-Hydraulic Stability," May 21, 1985.
770. " Report on June 7, 1975 Ferndale Earthquake," Pacific Gas & Electric Company, August 2, 1975.
771. Memorandum for W. Minners from L. Reiter, " Generic Issue No. B-50 Post Operating Basis Earthquake Inspection," June 7, 1985.
772. Letter to A. Schwencer (NRC) from C. Dunn (Duquesne Light Company), " Beaver Valley Power Station, Unit No. 1, Docket No. 50-334, Request for Amendment to the Operating License - No. 35," October 27, 1978.
773. Letter to J. Carey (Duquesne Light Company) from S. Varga (NRC), " Beaver Valley Unit No.1 - Operation With Two Out of Three Reactor Coolant Loops
- Safety Evaluation," July 20, 1984.
774. Memorandum for D. Eisenhut from D. Wigginton, "Closecut of MPA E-05; Westinghouse N-1 Loop Operation," January 11, 1985.
(/
06/30/86 R-49 NUREG-0933
Revision 3 775. Note to G. Lainas from R. Clark, " Statue of Single Loop Operation for BWRs," October 2, 1984.
776. Memorandum for R. Bernero from D. Eisenhut, "BWR Thermal-Hydraulic Stability Technical Specifications," November 16, 1984.
777. Memorandum for W. Dircks from H. Denton, "Closecut of TMI Action Plan Items I.A.2.2 and I.A.2.7 Training and Qualifications of Operating Personnel," June 24, 1985.
778. Memorandum for W. Dircks from H. Denton, "TMI Action Item I.A.3.4,"
February 12, 1985.
779. Memorandum for W. Dircks from J. Taylor, "TMI Action Plan - Completed Item," June 26, 1985.
780. IE Information Notice No. 83-58, "Transamerica DeLaval Diesel Generator Crankshaft Failure," U.S. Nuclear Regulatory Commission, August 30, 1983.
781. IE Information Notice No. 83-51, " Diesel Generator Events," d.S. Nuclear Regulatory Commission, August 5, 1983.
782. Memorandum for C. Berlinger from H. Denton, " Detail Assignment to D0L, Transamerica DeLaval Emergency Diesel Generator Project Group (TDI Project Group)," January 25, 1984.
783. SECY-84-34, " Emergency Diesel Generators Manufactured by Transamerica { )
DeLaval, Inc.," January 25, 1984.
784. Letter to D. Bixby (TDI) from D. Eisenhut (NRC), February 14, 1984.
785. TDI Diesel Generators Owners' Group Program Plan, March 2, 1984.
786. SECY-84-155, "Section 208 Report to the Congress on Abnormal Occurrences for October-December,1983," April 11,1984.
787. Letter to J. B. George (Transamerica Delaval, Inc., Owners' Group) from D. Eisenhut (NRC), " Safety Evaluation Report, Transamerica Delaval, Inc.
Diesel Generator Owners' Group Program Plan," August 13, 1984.
788. Memorandum for W. Minners from B. Sheron, " Additional Low-Temperature-Overpressure Protection Issues for Light-Water Reactors," August 1, 1984.
789. IE Information Notice No. 83-26, " Failure of Safety / Relief Valve Discharge Line Vacuum Breakers," U.S. Nuclear Regulatory Commission, May 3, 1983.
790. NUREG/CR-3384, " VISA - A Computer Code for Predicting the Probability of Reactor Pressure Vessel Failure," U.S. Nuclear Regulatory Commission, September 1983.
791. Memorandum for K. Seyfrit from C. Hsu, "EE No. AE0D/E322 Damage to Vacuum Breaker Valves as a Result of Relief Valve Lif ting," September 21, 1983.
l I 06/30/86 R-50 NUREG-0933
Revision 3 ys
( ) 792. AE0D/C401, " Low Temperature Overpressure Events at Turkey Point Unit 4,"
(s_,/ Office for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, March 1984.
793. Memorandum for B. Sheron from B. Liaw, " Additional Low-Temperature-Overpressure Protection (LTOP) Issues for Light-Water Reactors," August 30, 1984.
794. Memorandum for A. Seyfrit from E. Imbro, " Single Failure Vulnerability of Power Operated Relief Valve Actuation _ Circuitry for Low Temperature Overpressure Protection (LTOP)," October 24, 1984.
795. AE0D/C403, "Edwin I. Hatch Unit No. 2 Plant Systems Interaction Event on August 25, 1982," Office for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, May 1984.
796. Memorandum for R. Mattson from T. Dunning, "RHR Interlocks for Westinghouse Plants," April 17, 1984.
797. Memorandum for F. Rowsome from W. Houston,'"RCS/RHR Suction Line Valve Interlock on PWRs," August 27, 1984.
798. NSAC-52, " Residual Heat Removal Experience Review and Safety Analysis, Pressurized Water Reactor," Nuclear Safety Analysis Center, January 1983.
es 799. Memorandum for W. Dircks from H. Denton, " Resolution of Generic
- ) Issue III.D.2.3 -- Liquid Pathway Studies," August 28, 1985.
U 800. NUREG/CR-4258, "An Approach to Team Skills Training of Nuclear Power Plant Control Room Crews," U.S. Nuclear Regulatory Commission, July 1985.
801. Memorandum for W. Dircks from H. Denton, " Team Training for Nuclear Power Plant Control Room Crews," July 10, 1985.
802. NUREG/CR-3739, "The Operator Feedback Workshop: A Technique for Obtaining Feedback from Operations Personnel," U.S. Nuclear Regulatory Commission, September 1984.
803. NUREG/CR-4139, "The Mailed Survey: A Technique for Obtaining Feedback from Operations Personnel," U.S. Nuclear Regulatory Commission, May 1985.
804. Memorandum for W. Dircks from H. Denton, "TMI Action Plan Item I.A.2.6(4),"
September 25, 1985.
805. Memorandum for T. Combs from H. Denton " Revised SRF Section 13.5.2 and Appendix A to SRP Section 13.5.2 of NUREG-0800," July 17, 1985.
806. Memorandum for W. Dircks from H. Denton, " Closeout of TMI Action Plan, Task II.B.6, ' Risk Reduction for Operating Reactors at Sites With High Population Densities,'" September 25, 1985.
807. Memorandum for W. Dircks from R. Minogue, " Closeout of TMI Action Plan
((h
/ Task II.B.8 'Rulemaking Proceeding on Degraded Core Accidents - Hydrogen Control,'" July 19, 1985.
06/30/86 R-51 NUREG-0933 1
Rsvision 3 808. Memorandum for W. Dircks from H. Denton, "Close Out of TMI Action Plan, Task II.B.8," August 12, 1985.
809. NUREG-1070, "NRC Policy on Future Reactor Designs," U.S. Nuclear Regulatory Commission, July 1985.
810. NUREG/CR-3085, " Interim Reliability Evaluation Program: Analysis of.the Millstone Point Unit 1 Nuclear Power Plant," U.S. Nuciear Regulatory Com-mission, (Vol. 1) April 1983, (Vol. 2) August 1983, (Vol. 3) July 1983, (Vol. 4) July 1983.
811. NUREG/CR-3511, " Interim Reliability Evaluation Program: Analysis of the Calvert Cliffs Unit 1 Nuclear Power Plant," U.S. Nuclear Regulatory Commission, (Vol. 1) May 1984, (Vol. 2) October 1984.
812. NUREG/CR-2728, " Interim Reliability Evaluation Program Procedures Guide,"
U.S. Nuclear Regulatory Commission, March 1983.
813. Memorandum for W. Dircks from R. Minogue, "Closecut of TMI Action Plan, Task II.C.1, ' Interim Reliability Evaluation Program,'" July 9, 1985.
814. SECY-84-133, " Integrated Safety Assessment Program (ISAP)," March 23, 1984.
815. SECY-85-160, " Integrated Safety Assessment Program - Implementation Plan,"
May 6, 1985.
816. Memorandum for W. Dircks from H. Denton, "Close-out of Generic Issues II.C.2,
' Continuation of IREP,' and IV.E.5, ' Assess Currently Operating Reactors,'"
September 25, 1985.
817. Memorandum for W. Dircks from R. Minogue, "Closecut of TMI Action Plan Task II.E.2.2, Research on Small Break LOCA's and Anomalous Transients,'"
July 25, 1985.
818. Memorandum for W. Dircks from J.-Taylor, "TMI Action Plan - Completed Item," August 15, 1985.
819. EPRI EL-3209, " Workshop Proceedings: Retaining Rings for Electric Generators," Electric Power Research Institute, August 1983.
820. Memorandum for R. Fraley from R. Vollmer, " Proposed NRR Revisions to Review Procedures for Turbine Missile Issue," May 12, 1983.
821. Memorandum for W. Johnson from T. Novak, " Midland SSER #3 - Turbine Missile Review," November 1, 1983.
822. Memorandum for V. Stello from H. Denton, "NRR Plans for Approval of WCAP-10271," January 11, 1985.
823. Letter to J. Sheppard (Westinghouse Owners Group) from C. Thomas (NRC),
" Acceptance for Referencing of Licensing Topical Report WCAP-10271,
' Evaluation of Surveillance Frequercies and Out of Service Times for the Reactor Protection Instrumentation Systems,'" February 21, 1985.
06/30/86 R-52 NUREG-0933 1
e Revision 3
,m 824. Memorandum for T. Speis from R. Mattson, " Request for Prioritization of V)
(
Generic Safety Issue - Failure of HPCI Steam Line Without Isolation,"
October 18, 1983.
825. Memorandum for K. Seyfrit from P. Lam, " Failure of an Isolation Valve of the Reactor Core Isolation Cooling System to Open Against Operating Reactor Pressure," August 23, 1984.
826. Letter to A. Schwenser (NRC) from J. Kemper (Philadelphia Electric Company),
" Limerick Generating Station, Units 1 and 2, Request for Additional Infor-mation from NRC Equipment Qualification Branch (EQB)," February 27, 1984.
827. NED0-24708A, " Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors," General Electric Company, December 1980.
828. NUREG/CR-3933, " Risk Related Reliability Requirements for BWR Safety-Important Systems with Emphasis on the Residual Heat Removal System," U.S.
Nuclear Regulatory Commission, August 1984.
829.- BNL A-3740, "An Evaluation of Unisolated LOCA Outside the Drywell in the Shoreham Nuclear Power Station," Brookhaven National Laboratory, June 1985.
830. Memorandum for W. Minners from A. Thadani, " Comments on Generic Issue No. 87 - Failure of HPCI Steam Line Without Isolation," June 28, 1985.
p 831. NUREG/CR-1433, " Examination of the Use of Potassium Iodide (KI) as an
,' ((/ j Emergency Protective Measure for Nuclear Reactor Accidents," U.S. Nuclear Regulatory Commission, October 1980. -
832. SECY-83-362, " Emergency Planning - Predistribution/ Stockpiling.of Potassium Iodide for the General Public," August 30, 1983.
833. SECY-85-167, " Federal Policy Statement on the Distribution and Use of Potassium Iodide," May 13, 1985.
834. Memorandum for H. Denton and R. Minogue from W. Dircks, " Review of NRC Requirements for Nuclear Power Plant Piping," August 1, 1983.
835. Memorandum for W. Dircks from R. Minogue, " Plan to Implement Piping Review Committee Recommendations," July 30, 1985.
836. Memorandum for T. Murley, et al. , from J. Taylor, "Results of Regional Survey of Plant Specific Information Relating to the Potential for Uncon-trolled Radiation Exposures in PWR Reactor Cavities," June 18, 1985.
837.-Note to R. Vollmer from T. Speis, " Proposed Request to Perform Research on the Stress Corrosion Cracking of Pressure Boundary Ferritic Steels in Selected Environments," January 7,1985.
838. NUREG-1165, " Environmental Standard Review Plan for ES Section 7.1.1,"
U.S. Nuclear Regulatory Commission, November 1985.
s
) 839. Letter to J. Bayne (PASNY) from S. Varga (NRC), " Steam Generator Tube and Girth Weld Repairs at the Indian Point Nuclear Generating Plant, Unit No. 3 (IP-3)," May 27, 1983.
06/30/86 R-53 NUREG-0933
RQvision 3 840. "Value-Impact Analysis of Recommendations Concerning Steam Generator Tube Degradations and Rupture Events," Science Applications, Inc., February 2, 1983.
841. Regulatory Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commission (Rev. 1) April 1977.
842. IE Information Notice No. 82-37, " Cracking in the Upper Shell to Transition Cone Girth Weld of a Steam Generator at an Operating Pressurized Water Reactor," U.S. Nuclear Regulatory Commission, September 16, 1982.
843. Letter to D. Smith (NRC) from E. Rahe (Westinghouse), January 17, 1982.
844. NUREG/CR-3281, " Investigation of Shell Cracking on the Steam Generators at Indian Point Unit No. 3," U.S. Nuclear Regulatory Commission, June 1983.
845. NUREG/CR-3614, " Constant Extension Rate Testing of SA302 Grade B Material in Neutral and Chloride Solutions," U.S. Nuclear Regulatory Commission, February 1984.
846. Letter to W. Hazelton (NRC) From H. Watanabe (GE), "' Laboratory Examination of Garigliano Secondary Steam Generator-B Core Samples,' NEDE-25162, July 1979," December 13, 1979.
847. EPRI NP-1136, " Limiting Factor Analysis of High Availability Nuclear Plants (Boiling Water Reactors)," Electric Power Research Institute, (Vol. 1) l l August 1979.
848. Regulatory Guide 1.56, " Maintenance of Water Purity in Boiling Water Reactors," U.S. Nuclear Regulatory Commission, (Rev.1) July 1978.
849. NUREG/CR-3842, " Steam Generator Group Project Task 8 - Selective Tube Unplugging," U.S. Nuclear Regulator Commission, July 1984.
850. NRC Letter to All PWR Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits, and Ft. St. Vrain,
" Staff Recommended Actions Stemming from NRC Integrated Program for the Resolution of Unresolved Safety Issues Regarding Steam Generator Tube Integrity (Generic Letter 85-02)," April 17, 1985.
851. NUREG/CP-0058, "Twelf th Water Reactor Safety Research Information Meeting,"
U.S. Nuclear Regulatory Commission, (Vol. 4) January 1985.
852. NUREG/CP-0044, " Proceedings of the International Atomic Energy Agency Specialists' Meeting on Subcritical Crack Growth," U.S. Nuclear Regulatory Commission, (Vol. 1) May 1983, (Vol. 2) May 1983.
853. Scott, P. et al., " Corrosion Fatigue Crack Growth in Reactor Pressure Vessel Steels - Structural Integrity of Light Water Reactor Components,"
Elseviar Science Publishing Co., Inc., 1982.
O 06/30/86 R-54 NUREG-0933
Revision 3
[m 854. NUREG/CR-4121, "The Effects of Sulfur Chemistry and Flow Rate on Fatigue V) Crack Growth Rates in LWR Environments," U.S. Nuclear Regulatory Commission, February 1985.
855. NUREG-0975, " Compilation of Contract Research for the Materials Engineering Branch, Division of Engineering Technology," U.S. Nuclear Regulatory Com-mission, (Vol. 2) March 1984.
856~. PNO-II-85-41, "Small Steam Generator Surface Cracks," U.S. Nuclear Regulatory Commission, April 23, 1985.
857. Memorandum for W. Minners from B. Liaw, "Prioritization of Generic Issue No. (111) Stress Corrosion Cracking of RCPB Ferritic Steels and Steam Generator Vessels," June 7, 1985.
858. IE Information Notice No. 85-65, " Crack Growth in Steam Generator Girth Welds," U.S. Nuclear Regulatory Commission, July 31, 1985.
859. Memorandum for H. Thompson from J. Knight, " Steam Generator Shell Transition Joint Cracking," July 10, 1985.
860. NUREG-0937, " Evaluation of PWR Response to Main.Steamline Break With Con-current Steam Generator Tube Rupture and Small-Break LOCA," U.S. Nuclear Regulatory Commission, December 1982.
861. SECY-83-357B, " Status of Hydrogen Control Issue and Rulemaking Recommenda-g ) tions in SECY-83-357A," December 3, 1984.
J 862. IE Bulletin No. 79-13, " Cracking in Feedwater System Piping," June 25, 1979.
863. Memorandum for T. Speis from H. Denton, " Closeout of Generic Issues B-58 and C-11," July 9, 1985.
864. AE0D/C301, " Failures of Class 1E Safety-Related Switchgear Circuit Breakers to Close on Demand," Office for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, April 1983.
865. Memorandum for T. Speis from H. Denton, " Resolution of Generic Issue 14,
'PWR Pipe Cracks,'" October 4, 1985.
866. Federal Register Notice 47 FR 7023, " Proposed Policy Statement on Safety Goals for Nuclear Power Plants," February 17, 1982.
867. Federal Register Notice 48 FR 10772, " Safety Goal Development Program,"
March 14, 1983.
868. Letter to J. Ahearne from M. Plesset, " Recommendations of President's Commission on ACRS Role," January 15, 1980.
869. Federal Register Notice 46 FR 22358, "10 CFR Part 2, ACRS Participation O in NRC Rulemaking," April 17, 1981.
N.]
06/30/86 R-55 NUREG-0933
Revision 3 870. Memorandum for Commissioner Ahearne, et al. , from L. Bickwit, Jr. , et al., "TMI Action Plan, Chapter V, Formal Procedures for Ensuring Periodic Public Interaction," October 2, 1980.
871. Memorandum for W. Dircks, et al., from J. Hoyle, " Staff Requirements -
Discussion of Action Plan, Chapter V (See SECY-80-2308), 2:00 p.m. Monday, July 7, 1980, Commissioners' Conference Room, D.C. Office (0 pen to Public Attendance)," July 9, 1980.
872. Federal Register Notice 45 FR 49535, "10 CFR Part 2, Procedural Assistance in Adjudicatory Licensing Proceedings," July 25, 1980.
873. Federal Register Notice 46 FR 13681, "10 CFR Part 2, Domestic Licensing Proceedings; Procedural Assistance Program," February 24, 1981.
874. Memorandum for L. Bickwit from S. Chilk, "SECY-81-391 - Provision of Free Transcripts to All Full Participants in Adjudicatory Proceedings: May 11, 1981 Comptroller General Decision," February 25, 1982.
875. Federal Register Notice 45 FR 34279, "10 CFR Parts 2, 50, Possible Amend-ments to 'Immediate Effectiveness Rule,'" May 22, 1980.
876. Federal Register Notice 47 FR 47260, "10 CFR Part 2, Commission Review Procedures for Power Reactor Construction Permits; Immediate Effectiveness Rule," October 25, 1982.
877. Federal Register Notice 48 FR 50550, "10 CFR Part 2, Rules of Practice for Domestic Licensing Proceedings; Role of NRC Staff in Adjudicatory Licensing Hearings," November 2, 1983.
878. NUREG-0632, "NRC Views and Analysis of the Recommendations of the Presi-dent's Commission on the Accident at Three Mile Island," U.S. Nuclear Regulatory Commission, November 1979.
879. Federal Register Notice 46 FR 28533, " Statement of Policy on Conduct of Licensing Proceedings," May 27, 1981.
880. Memorandum to All Employees from N. Palladino, " Regulatory Reform Task Force," November 17, 1981.
881. Letter to the Honorable Thomas P. O'Neill, Jr. from N. Palladino, February 21, 1983.
882. Federal Register Notice 48 FR 44173, "10 CFR Part 50, Revision of Back-fitting Process for Power Reactors," September 28, 1983.
883. Federal Register Notice 48 FR 44217, "10 CFR Part 50, Revision of Backfit-ting Process for Power Reactors," Septembe' 28, 1983.
884. Federal Register Notice b0 FR 38097, "10 CFR Parts 2 and 50, Revision of Backfitting Process for Power Reactors," September 20, 1985.
O 06/30/86 R-56 NUREG-0933
Revision 3 O 885. Memorandum for H. Thompson from D. Crutchfield, " Potential Immediate (Q Generic Actions as a Result of the Davis-Besse Event of June 9, 1985,"
August 5, 1985.
886. NUREG-1154, " Loss.'of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9,1985," U.S. Nuclear Regulatory Commission, July 1985.
887. Memorandum for T. Speis from H. Thompson, "Short Term Generic Actions as a Result of the Davis-Besse Event of June.9, 1985," August 19, 1985.
888. Memorandum for H. Denton from T. Speis, " Adequacy of the Auxiliary Feedwater System at Davis-Besse," July 23, 1985.
889. NSAC-60, "A Probabilistic Risk Assessment of Oconee Unit 3," Electric Power Research Institute, June 1984.
890. NUREG-1032, " Evaluation of Station Blackout Accidents at Nuclear Power Plants," U.S. Nuclear Regulatory Commission, (Draf t) May 1985.
891. Letter to T. Novak (NRC) from R. Crouse (Toledo Edison Company),
December 31, 1981.
892. NUREG/CR-2770, " Common Cause Fault Rates for Valves," U.S. Nuclear Regula-tory Commission, February 1983.
893. NUREG/CR-2098, " Common Cause Fault Rates for Pumps," U.S. Nuclear Regulatory Commission, February 1983.
%./
894. Memorandum for 0. Parr from A. Thadani, " Auxiliary Feedwater System -
CRGR Package," November 9, 1984.
895. Memorandum for H. Denton, et al., from W. Dircks, " Staff Actions Resulting from the Investigation of the June 9 Davis-Besse Event (NUREG-1154),"
August 5, 1985.
896. SECY-86-56, " Status of Staff Study to Determine if PORVs Should be Safety Grade," February 18, 1986.
897. Memorandum for G. Lainas from F. Rowsome, " Safety Evaluation of the CE Licensees' Responses to TMI Action Item II.K.3.2," August 26, 1983.
898. Memorandum for G. Lainas from F. Rowsome, " Safety Evaluation of the B&W Licensees' Responses to TMI Action Item II.K.3.2," August 24, 1983.
899. Memorandum for G. Lainas from F. Rowsome " Safety Evaluation of the Westing-house Licensees' Responses to TMI Action Item II.K.3.2," July 22, 1983.
900. Memorandum for H. Thompson from W. Russell, " Comments on Draft List of Longer Term Generic Actions as a Result of the Davis-Besse Event of June 9, 1985," September 19, 1985.
g 901. Memorandum for T. Combs from H. Denton, " Revised SRP Section 9.2.1 and
) SRP Section 9.2.2 of NUREG-0800," June 24, 1986.
06/30/86 R-57 NUREG-0933
R:visicn 3 902. Memorandum for J. Sniezek and R. Fraley from H. Denton, " Resolution of { l Generic Issue No. 36, ' Loss of Service Water,'" May 13, 1986.
903. Memorandum for T. Speis from H. Denton, " Resolution of Generic Issue 3,
'Setpoint Drift in Instrumentation,'" May 19, 1986.
904. SECY-83-293, " Amendments to 10 CFR 50 Related to Anticipated Transients Without Scram (ATWS) Events," July 9, 1983.
905. Memorandum for T. Speis from R. Bernero, " Enhancement of the Reliability of Westinghouse Solid State Protection System (SSPS)," April 5, 1985.
906. NUREG/CR-3971, "A Handbook for Cost Estimating," U.S. Nuclear Regulatory Commission, October 1984.
907. Memorandum for W. Minners from B. Sheron, " Generic Issues C-4, C-5, C-6,"
May 29, 1985.
908. SECY-83-472, " Emergency Core Cooling System Analysis Methods," November 17, 1983.
909. AE0D/C503, " Decay Heat Removal Problems at U.S. Pressurized Water Reactors,"
Office for Analysis and Ev_aluation of Operational Data, U.S. Nuclear Regula-tory Commission, December 1985.
910. Memorandum for H. Denton from C. Heltemes, " Case Study Report - Decay Heat Removal Problems at U.S. Pressurized Water Reactors," December 23, 1985. { }
911. Memorandum for C. Heltemes from H. Denton, "AE0D's Report on Decay Heat Removal Problems at U.S. PWRs," February 10, 1986.
912. Memorandum to T. Murley, et al. , from H. Denton, " Evaluation of Industry Success in Achieving ALARA-Integrated Radiation Protection Plans - Data Trend Assessments," May 19, 1986.
913. Memorandum for V. Stello from H. Denton, " Resolution of Generic Issue III.D.3.1, ' Radiation Protection Plans,'" May 19, 1986.
914. Memorandum for H. Thompson and T. Speis from R. Bernero, " Request for Com-ments on Draft CRGR Package with Requirements for Upgrading Auxiliary Feed-water Systems in Certain Operating Plants," October 3, 1985.
915. Memorandum for W. Minners from A. Thadani, " Seismic Induced Relay Chatter Issue," March 22, 1985.
916.. Regulatory Guide 1.29, " Seismic Design Classification," U.S. Nuclear Regulatory Commission, June 1972, (Rev. 1) August 1973, (Rev. 2)
February 1976, (Rev. 3) September 1978.
917. Regulatory Guide 1.100, " Seismic Qualification of Electrical Equipment for Nuclear Power Plants," U.S. Nuclear Regulatory Commission, March 1976, (Rev. 1) August 1977.
06/30/86 R-58 NUREG-0933 l
Revision 3 918. NUREG/CP-0070, " Proceeding of the Workshop on Seismic and Dynamic Fragility 1 ..
of Nuclear Power Plant Components," U.S. Nuclear Regulatory Commission, August 1985.
919. NUREG-1030, " Seismic Qualification of Equipment in Operating Nuclear Power Plants," U.S. Nuclear Regulatory Commission, (Draft) August 1985.
920. ANSI /ANS 5.1, " Decay Heat Power in Light Water Reacters," American National Standards Institute, 1979.
4 l
l 1
l l
06/30/86 R-59 NUREG-0933
NnC PonM 335 U S. NUCLES.Z K EGULATOR Y COMMISSION i R t roR T NUM8 E R # Augaed &r TtOC ead vor N o. .r emys
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.C ,,a fiUREG-0933 mi. m2 ' BIBLIOGRAPHIC DATA SHEET Supplement 5 SEE INSTRUCTsON5 ON THE REVERSE isf LE ANO suBisTLE 3 LE AVE BL ANK l l A Prioritization of Generic Safety Issues 4 D ATE REPORT COMPLETED MONTM VEAR
. UTROms, June 1986 R. Emrit, H. VanderMolen, J. Pittman, W. Milstead, * "
e oArt aEPOar issuEo R. Riggs September l 1986 N free ld Gode $ PROJECTIT ASK/WORet uNeT NUM8ER Division o 7 v tR5,ORMINCg ORGANIEIf dalAME etyANhMAIL[NG eview an AODREff ftUversight Office of fluclear Reactor Regulation ,,,NcRGRANrNUM,ER U.S. Nuclear Regulatory Commission Washington, D.C. 20555
- 10. SPONSORtNG ORGANt2 ATION N AME ANO M AILING ADDRESS Itackde le Codel its T vPE OF REPORY Same as above . PE a,Oo COv E a to <,~~...e d.-
12 SUPPLEMENT ARY NOTES 93 in STH ACE (200 worde oe seul The report presents the priority rankingr for generic safety issues related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient l ) allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The safety priority rankings are HIGli, 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 resolutions of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative.
,. oOCuME Nr AN AL v 5.s . . .E voORos.oE sCR.. roms
,,AAy,3rv generic safety issues nuclear power plants Unlimited 16 SECURITv CL AS$is tC Af SON
<r. ,,,,e e soE Niisit asfoeEN E Not o TE aMs gjnrjanqjfjpg a rn,s reoorrr l I linc l a c c i f i nci Il NVV0tROFPAGE5 t H PRi( f