ML20211B067
| ML20211B067 | |
| Person / Time | |
|---|---|
| Issue date: | 05/28/1986 |
| From: | Lainas G Office of Nuclear Reactor Regulation |
| To: | Kniel K Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20209C518 | List:
|
| References | |
| FOIA-86-586, RTR-NUREG-0660, RTR-NUREG-660, TASK-***, TASK-TM NUDOCS 8606110465 | |
| Download: ML20211B067 (15) | |
Text
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
- k MAY 2 81996 MEMORANDUM FOR
- Kar'l Kniel, Chief Safety Program Evaluation Branch Division of Safety Review and Oversight FROM:
Gus Lainas, Assistant Director for BWR Division of BWR Licensing
SUBJECT:
REVIEW 0F NUREG-0660 ISSUES IN LIGHT OF THE CHERN0BYL ACCIDENT In response to the May 23, 1986 memo from H. R. Denton to the Division Directors and Director, ORAS, Enclosure 1 is a marked-up copy of the NUREG-0660 items identifying our suggestions for reconsideration. clarifies the reasons for selecting some of the items.
We also note some additional items for possible consideration when we get a better understanding of the Chernobyl event as follows:
emergency HVAC, respirators)gn features (i.e., shielding, 1.
Reconsider Control Room desi to reflect severe accident conditions (e.g., containment venting or failure).
2.
Reexamine remote shutdown panel design requirements with respect to LOCA conditions, including severe accident sequences.
3.
Reexamine the consequences of venting combustible gases (e.g., hydrogen) into the secondary containment. Consider the feasibility of providing an option for venting directly into the atmosphere to preclude secondary containment fires in the event of a severe accident sequence. Consider the performance of fire protection systems with respect to the above option.
4.
In view of the possibility that the Chernobyl event involved a fire in the turbine building, reexamine the storage and use of highly flammable combustible gases (e.g., hydrogen) at existing nuclear power plants.
5.
Management of resources for minimizing the impact of core melt / containment breach accidents.
Planning should include the marshalling of off-site resources, where needed, to supply needed materials and expertise.
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1 6.
Current technical specifications allow BWR plant start-up and shutdown with de-inerted, or partially de-inerted, containments.
This is done to minimize down-time due to maintenance.
If I
maintenance activities increase the likelihood of accidents, the de-inerting would concomitantly increase the vulnerability.
The impact on risk needs to be assessed.
7.
Traditional regulatory cost / benefit analyses do not consider property damage, or the cost of replacement power, following I
an accident. This weighs the cost / benefit studies against many potentially desirable plant improvements. This should be corrected.
[-
us C. Lainas, Assistant Director for BWR Division of BWR Licensing
Enclosures:
As stated cc w/ enclosures:
R. Bernero R. Houston B. D. Liaw W. Hodges M. Srinivasan D. Vassallo J. Kudrick K. Campe CONTACT:
L. Hulman, x-27941
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LIST OF POTENTIAL GENERIC ISSUES TO BE CONSIDERED I
FOLLOWING IHL CHLRNOBYL ACCIDENI Legend i
NOTES:
3(b) - Resolution Resulted in No New Requirements 4 - Issue to be Prioritized in the Future LOW
- Low Safety Priority DROP
- Issue Dropped as a Generic Issue I
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Action Lead lead Office /
Safety Plan Item /
SPEB Division /
Priorite-Issue No.
Title Engineer Branch Ranking' TMI ACTION PLAN ITEMS s.-
I.A OPERATING PERSONNEL f a 1' T-eialag e=d QuaLLEisatient ef Operat!ng Per :---!
I.A.2.2 Training and Qualifications of Operations Personnel Colmar NRR/DHFS/LQB NOTE 3(b) 1.A.2.5 Plant Drills Colmar NRR/DHFS/LQB NOTE 3(b) 1.A.2.6 Long-Term Upgrading of Training and Qualifications I.A.2.6(2)
Staff Review of HRR 80-117 Colmar NRR/DHFS/LQB NOTE 3(b)
I.A.2.6(4)
Operator Workshops Colmar NRR/DHFS/LQB NOTE 3(b) 1.A.2.6(5)
Develop Inspection Procedures for Training Program Colmar NRR/DHFS/LQB NOTE 3(b)
I.A.2.6(6)
Nuclear Power Fundamentals Colmar NRR/DHFS/LQB DROP I.A.2.7 i / Accreditation of Training Institutions Colmar NRR/DHFS/LQB NOTE 3(b)
I.A.3 Licensing and Requalification of Operating Personnel I.A.3.2 Operator Licensing Program Changes Emrit NRR/DHFS/0LB NOTE 3(b) 1.A.3.4 Licensing of Additional Operations Personnel Thatcher NRR/DHFS/LQB NOTE 3(b)
'^'
Si _'ste
'Jre red Oe"c'~;---t I.A.4.1 Initial Simulator Improvement I.A.4.1(1)
Short-Term Study of Training Simulators Thatcher NRR/DHFS/0LB NOTE 3(b)
I.A.4.2 9 / Long-Term Training Simulator Upgrade I.C OPERATING PROCEDURES I.C.1 Short-Term Accident Analysis and Procedures ~ Revision I.C.1(4)
Confirmatory Analyses of Selected Transients Riggs NRR/DSI/RSB NOTE 3(b) 2
l Action Lead lead Office /
Safety Plan Item /
SPEB Division /
Priority Issue No.
Title Engineer Branch Ranking 1.0 CONTROL ROOM DESIGN I.D.3 Safety System Status Monitoring Thatcher NRR/DHFS/HFEB MEDIUM I.D.5 Improved Control Room Instrumentation Research I.D.5(1)
Operator-Process Communication Thatcher RES/DF0/HFBR NOTE 3(b)
{
I.D.5(4)
Process Monitoring Instrumentation Thatcher RES/DF0/ICBR NOTE 3(b) i 1.F QUALITY ASSURANCE I.F.2 Develop More Detailed QA Criteria I.F.2(1)
Assure the Independence of the Organization Performing Pittman OIE/DQASIP/QUAB LOW the Checking Function I.F.2(4)
Establish Criteria for Determining QA Requirements Pittman OIE/DQASIP/QUAB LOW for Specific Classes of Equipment I.F.2(5)
Establish Qualification Requirements for QA and QC Pittman OIE/DQASIP/QUAB LOW Personnel I.F.2(7)
Clarify that the QA Program Is a Condition of the Pittman OIE/DQASIP/QUAB LOW Construction Permit and Operating License I
I.F.2(8)
Compare NRC QA Requirements with Those of Other Pittman OIE/DQASIP/QUAB LOW i
Agencies I.F.2(10)
Clarify Requirements for Maintenance of "As-Built" Pittman OIE/DQASIP/QUAB LOW l
Documentation l,
I.F.2(11)
Define Role of QA'in Design and Analysis Activities Pittman OIE/DQASIP/QUAB LOW 1
i II.A SITING i
!!.O.1 Siting PGIiC) Eefe.LelobiGG Y ' 50I C T.
NER/EE/ SAC NOIE 3(v)
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3
Action Lead lead Office /
- a ety Plan Item /
SPEB Division /
I r im it Issue No.
Title Engineer Branch taaking RELIABILITYENGINEERINGANDRISKASSESSHg II.C II.C.1 Interim Reliability Evaluation Program Pittman RES/DRA0/RRB NOTE 3(b)
J.C.2 ntinuation of Interim Reljability Evaluation Program Pittman NRR/ DST /,RRAB NOTE 3(b) m
,=.f4 C.3 NC hW4 c.4 i o r1 II.D R
TOR COOLANT SYSTEM RELIEF AND SAFETY VALVES II.D.2 Research on Relief and Safety Valve Test Requirements Riggs RES LOW II.E SYSTEM DESIGN II.E.2 Emergency Core Cooling System II.E.2.2 Research on Small Break LOCAs and Anomalous Transients Riggs RES/DAE/RSRB NOTE 3(b)
II.E.2.3 Uncertainties in Performance Predictions V'Molen NRR/DSI/RSB LOW II.E.3 Decay Heat Removal II.E.3.4 Alternate Concepts Research Riggs RES/DAE/FBRB NOTE 3(b)
II.E.4 Containment Design II.E.4.4 Purging Milstead NRR/DSI/CSB NOTE 3(b)
II.E.4.4(4)
Evaluate Purging and Venting During Normal Operation II.E.4.4(5)
Issue Modified Purging and Venting Requirement Milstead NRR/DSI/CSB NOTE 3(b)
II.F INSTRUMENTATION AND CONTROLS II.F.4 Study of Control and Protective Action Design Thatcher NRR/DSI/ICSB DROP Requirements 4
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l Lead Lead Office /
Safety Action SPEB Division /
Priority Plan Item /
Issue No.
Title Engineer Branch Ranking II.H TMI-2 CLEANUP AND EXAMINATION II.H.1 Maintain Safety of TMI-2 and Minimize Environmental Matthews NRR/TMIPO NOTE 3(b)
Impact III.A EMERGENCY PREPAREDNESS AND RADIATION EFFECTS III.A.1 Improve Licensee Emergency Preparedness - Short Term i
!!! ^.1.2 M:intsia C;pplic ci-Thyroid "lecking agcat 1
III.A.1.3(1)
Workers Riggs OIE/DEPER/EPB NOTE 3(b)
III.A.1.3(2)
Public Riggs OIE/DEPER/EPB NOTE 3(b) i III.A.3 Improving NRC Emergency Preparedness III.A.3.1 NRC Role in Responding to Nuclear Emergencies III.A.3.1(1)
Define NRC Role in Emergency Situations Riggs OIE/DEPER/IRDB NOTE 3(b)
III.A.3.1(2)
Revise and Upgrade Plans and Procedures for the NRC Riggs OIE/DEPER/IRDB NOTE 3(b)
Emergency Operations Center III.A.3.1(3)
Revise Manual Chapter 0502, Other Agency Procedures, Riggs OIE/DEPER/IRDB NOTE 3(b) and NUREG-0610 III.A.3.1(4)
Prepare Commission Paper Riggs DIE /DEPER/IRDB NOTE 3(b)
III.A.3.1(5)
Revise Implementing Procedures and Instructions for Riggs OIE/DEPER/IRDB NOTE 3(b)
Regional Offices Riggs 01E/DEPER/IRDB NOTE 3(b)
III.A.3.2 Improve Operations Centers Thatcher OIE/DEPER/IRDB NOTE 3(b) l III.A.3.4 Nuclear Data Link Pittman OIE/DEPER/IRDB NOTE 3(b)
III.A.3.5 Training, Drills, and Tests III.A.3.6 Interaction of HRC and Other Agencies Pitt- ;
DIE /DEPER/EPLB NOTE 3(b) t 111 a 1 c(1) f atarnational-
-- - - - --- - - P i t tman
-DIE /DEPER/EPLB NOTE 3(b) r dera!
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- Pitt --
O!E/DEPER/EPLB NOTE 3(b) e
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Action lead lead Office /
Tafety Plan Item /
SPEB Division /
Priority Issue No.
Title Engineer Branch Ranking III.B EMERGENCY PREPAREDNESS OF STATE AND LOCA'. GOVERNMENTS
! ". ".1 Tr ::ftr of " :;;;a+ibi+ibies--to-FEMA -
MMstead OIE/DEPER/IRDB NOTE 3(b) 111." 2 I plementation-of-NR M FEMA-Responsibilitie:
!!! ":2(1)
Me-kisens i ng-Proces s---
Miletead OIE/DEPER/IRDB NOTE 3(b)
!!!.".2(2)
Federal-Guidance-----
M Hsteed-OIE/DEPER/IRDB NOTE 3(b) 111.0 RADIATION PROTECTION III.D.1 Radiation Source Control III.D.1.1 Primary Coolant Sources Outside the Containment Structure l
III.D.1.1(2)
Review Information on Provisions for Leak Detection Emrit NRR/DSI/METB NOTE 4 j
III.D.1.1(3)
Develop Proposed System Acceptance Criteria Emrit NRR/DSI/METB NOTE 4 I11.D.1.2 Radioactive Gas Management Emrit NRR/DSI/METB DROP l
_!! ".1.3 Y::t? ?:ti:n Sysk: er.d ":dio4edine-Adserbee-Criteri
'" " 1. 2(1)
": id; i";;6her-4-icenseet-Showid-Perform-Studie: --i E:rit NRR/DSI/METB DROP Make Modifications
!!!.".1.?(2)
" vie.: cad "evise--SRP Earit NRR/DSI/METB DROP C;- it NRR/DSI/METB DROP m.".1.0(3)
"a g re Licer;;;; te-Si, grade Eiltration-Systea; c
III.D.l.3(4)
Sponsor Studies to Evaluate Charcoal Adsorber Emrit NRR/DSI/METB NOTE 3(b) c " stem-Des ign-Fea tu re s -to-A id-in-Aceiden t Errit NRR/DSI/METB DROP
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De nugyy 3nd gggant 3 min 3((33 III.D.2 Public Radiation Protection Improvement III.D.2.1 Radiological Monitoring of Effluents III.D.2.1(1)
Evaluate the Feasibility and Perform a Value-Impact Emrit NRR/DSI/METB LOW Analysis of Modifying Effluent-Monitoring Design Criteria 6
- a Action Lead Lead Office /
Safety Plan Item /
SPEB Division /
Priority Issue No.
Title Engineer Branch Ranking
!!!.".2.h(2)
Stdj the EeasibHity-of-Requiring-the Development--
Emrit
"""/"SI/MEE LGW --
cf Effeetiv "
.s fer-Monitoring-and SampHng-Notrie-C=cae mad D=diojodine. Released to-the-Atmospi m III.D.2.1(3)
Revise Regulatory Guides Emrit NRR/DSI/METB LOW III.D.2.2 Radioiodine, Carbon-14, and Tritium Pathway Dose Analysis III.D.2.2(1)
Perform Study of Radioiodine, Carbon-14, and Tritium Emrit NRR/DSI/RAB NOTE 3(b)
Behavior III.D.2.3 Liquid Pathway Radiological Control III.D.2.3(1)
Develop Procedures to Discriminate Between Emrit NRR/DE/EHEB NOTE 3(b)
Sites / Plants III.D.2.3(2)
Discriminate Between Sites and Plants That Require Emrit NRR/DE/EHEB NOTE 3(b)
Consideration of liquid Pathway Interdiction Techniques
- !!! ".2.?'?)
E:td! 9 rerib!: M-tSed ef-Petbay Interdict-ion "GTE 3GJ -
r' III.D.2.3(4)
Prepare a Summary Assessment Emrit NRR/DE/EHEB NOTE 3(b)
III.D.2.4 Offsite Dose Measurements III.D.2.4(1)
Study Feasibility of Environmental Monitors V'Molen NRR/DSI/RAB NOTE 3(b)
III.9.2.5 Of frit: "::: C:ltuletion-Menuel V' Meter,
%""/GS;/^AC NOTE 3Cu)
IV.C EXTEND LESSONS LEARNED TO LICENSED ACTIVITIES OTHER THAN POWER REAC10RS IV.C.1 Extend lessons Learned from TMI to Other NRC Programs Emrit NMSS/WM NOTE 3(b)
IV.E SAFETY DECISION-MAKING IV.E.5 Assess Currently Operating Reactors Matthews NRR/DL/SEPB NOTE 3(b) 7
Action Lead Lead Office /
Safety Plan Item /
SPEB Division /
Priorit" Issue No.
Title Engineer Branch Ranking' IV.F FINANCIAL DISINCENTIVES TO SAFETY s.
4 IV.F.1 Increased OIE Scrutiny of the Power-Ascension Test Thatcher 01E/DQASIP NOTE 3(b)
Program IV. F. 2-Evaluate the Impacts of Financial Disincentives to Matthews SP NOTE 3(b) the Safety of Nuclear Power Plants i
NEW GENERIC ISSUES 1.
Failures in Air-Monitoring, Air-Cleaning, and Emrit NRR/DSI/METB DROP Ventilating Systems i
2.
Failure of Protective Devices on Essential Equipment Colmar NRR/DSI/ICSB NOTE 4 4.
End-of-Life and Maintenance Criteria Thatcher NRR/DE/EQB NOTE 3(b) 6.
Separation of Control Rod from Its Drive and BWR High V'Molen NRR/DSI/CPB NOTE 3(b)
Rod Worth Events 7.
Failures Due to Flow-Induced Vibrations V'Molen NRR/DSI/RSB DROP 10.
Surveillance and Maintenance of TIP Isolation Valves Riggs NRR/DSI/ICSB DROP and Squib Charges 12.
BWR Jet Pump Integrity Sege NRR/DE/MTEB, NOTE 3(b)
MEB 13.
Small Break LOCA from Extended Overheating of Riani NRR/DSI/RSB DROP Pressurizer Heaters l
14.
PWR Pipe Cracks Emrit NRR/DE/MTEB NOTE 3(b) 15.
Radiation Effects on Reactor Vessel Supports Emrit NRR/DE/MTEB LOW 17.
Loss of Offsite Power Subsequent to LOCA Colmar NRR/DSI/PSB, DROP ICSB 20 Effects of Electromagnetic Pulse on Nuclear Plant Thatcher NRR/DSI/ICSB NOTE 3(b)
Systems 21.
Vibration Qualification of Equipment Thatcher NRR/DE/EQB NOTE 4 22.
Inadvertent Boron Dilution Events V'Molen NRR/DSI/RSB NOTE 3(b) 8
o Action lead lead Office /
Safety Plan Item /
SPEB Division /
Priority Issue No.
Title Engineer Branch Ranking i
24.
Automatic Emergency Core Cooling System Switch to V'Molen NRR/DSI/RSB NOTE 4 Recirculation i
30.
Potential Generator Missiles - Generator Rotor Pittman NRR/DE/MEB DROP Retaining Rings 34.
RCS Leak Riggs NRR/DHFS/PSRB DROP 35.
Degradation of Internal Appurtenances in LWRs V'Molen NRR/DSI/CPB, LOW i
RSB 38.
Potential Recirculation System Failure as a Consequence Milstead NRR NOTE 4 i
of Injection of Containment Paint Flakes or Other Fine Debris 43.
Contamination of Instrument Air Lines Milstead NRR/DSI/ASB DROP 47.
Loss of Off-Site Power Thatcher NRR/DSI/RSB, NOTE 3(b)
ASB i
50.
Reactor Vessel level Instrumentation in BWRs Thatcher NRR/DSI/RSB, NOTE 3(b)
ICSB 53.
Consequences of a Postulated Flow Blockage Incident V'Molen NRR/DSI/CPB, DROP in a BWR RSB 55.
Failure of Class 1E Safety-Related Switchgear Circuit Emrit NRR/DSI/PSB DROP Breakers to Close on Demand 57.
Effects of Fire Protection System Actuation V'Molen NRR NOTE 4 on Safety-Related Equipment 58.
Inadvertent Containment Flooding Sege NRR/DSI/ASB, DROP CSB 62.
Reactor Systems Bolting Applications V'Molen NRR NOTE 4 63.
Use of Equipment Not Classified as Essential to Safety V'Molen NRR NOTE 4 in BWR Transient Analysis l
64.
Identification of Protection System Instrument Sensing Thatcher NRR/DSI/ICSB NOTE 3(b)
Lines 67.
Steam Generator Staff Actions 67.5.3 Secondary System isolation Riggs NRR/DSI/RSB DROP l
9 4
Action Lead Lead Office /
ta'iiy Plan Item /
SPEH Division /
I r w iit-Issue No.
Title Engineer Branch Eanke g' n
69.
Make-up Hozzle Cracking in B&W Plants Colmar NRR/DE/MEB, h01E 3(b)
MTEB 71.
Failure of Resin Demineralizer Systems and Their Emrit NRR NOTE 4 Effects on Nuclear Power Plant Safety 72.
Control Rod Drive Guide Tube Support Pin Failures V'Molen NRR NOTE 4 73.
Detached Thermal Sleeves Colmar NRR NOTE 4 74.
Reactor Coolant Activity limits for Operating Reactors Milstead NRR NOTE 4 76.
Instrumentation and Control Power Interactions Colmar NRR NOTE 4
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78.
Monitoring of Fatigue Transient limits for Reactor Riggs NRR NOTE 4 Coolant System 80.
Pipe Break Effects on Control Rod Drive Hydraulic Lines V'Molen NRR/DSI/RSB, LOW in the Drywells of BWR Mark I and II Containments
- ASB, i
CPB SL h;;;t of L :ked-Doors-end-Barriers
. Plante-Personne' C ' cr NRR/BHFS/PSRO O R0"--
sma cor-+o S2.
l..t-el-R'oea ll.L-;tebili ty Matthc-e N;; -
NOTE 4 85.
Reliability of Vacuum Breakers Connected to Steam Milstead NRR/DSI/CSB DROP Discharge Lines Inside BWR Containments 88.
Earthquakes and Emergency Planning Emrit NRR NOTE 4 89.
Stiff Pipe Clamps Riggs NRR NOTE 4 90.
Technical Specifications for Anticipatory Trips V'Molen NRR/DSI/RSB, LOW ICSB 92.
Fuel Crumbling During LOCA V'Molen NRR/DSI/RSB, LOW CPB 95.
Loss of Effective Volume for Containment Recirculation Milstead NRR NOTE 4 Spray 96.
RHR Suction Valve Testing V'Molen NRR NOTE 4 98.
CR0 Accumulator Check Valve Leakage Pittman NRR/DSI/ASB DROP 100.
OTSG Level Riggs NRR NOTE 4 104.
Reduction of Boron Dilution Requirements V'Holen NRR NOTE 4 106.
Piping and Use of Highly Combustible Gases in Vital Colmar NRR NOTE 4 Areas 107.
Generic Implications of Main Transformer Failures Colmar NRR NOTE 4 109.
Reactor Vessel Closure Failure V'Molen NRR NOTE 4 10 1
I Action Lead Lead Office /
Safety Plan Item /
SPEB Division /
Priority Issue No.
Title Engineer Branch Ranking 110.
Equipment Protective Devices on Engineered Safety Milstead NRR NOTE 4 Features 113.
Dynamic Qualification Testing of Large Bore Riggs NRR NOTE 4 Hydraulic Snubbers 114.
Seismic-Induced Relay Chatter Pittman NRR NOTE 4 115.
Reliability of Westinghouse Solid State Milstead NRR NOTE 4 Protection System 116.
Accident Management Pittman NRR/DHFS NOTE 4 117.
Allowable Outage Times for Diverse Simultaneous NRR NOTE 4 Equipment Outages 118.
Tendon Anchorage Failure Milstead NRR NOTE 4' 120.
On-Line Testability of Protection Systems Milstead NRR NOTE 4 122.
Davis-Besse Loss of All Feedwater Event of June 9, 1985 - Short-Term Actions 122.3 Physical Security System Constraints V'Molen NRR NOTE 4 n ric:g r:gg :
+ u og7_.htiens Governi~; 09^ 2-J De= =
- 40TE """
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Ein;;!:<:ilur: Criteris Swigested-by-the-Davis "c;;c Efr-t f Jun: 9, 2005-125.
Davis-Besse loss of All Feedwater Event of V'Molen NRR NOTE 4 June 9, 1985 - Long-Term Actions 125.1.1 Availability of the STA V'Molen NOTE 4 125.I.2 PORV Reliability Weinkam NOTE 4 NOTE 125.I.2.a Test Program Weinkam 125.1.2.b Surveillance Weinkam NOTE 4 125.I.2.c Auto Block Valve Closure Weinkam NOTE 4 125.I.2.d Equipment Qualification for Deed-and-Bleed Weinkam NOTE 4 Environment 125.I.3 SPDS Availability Weinkam NOTE 4 125.I.4 Plant-Specific Simulator V'Molen NOTE 4 125.I.5 Safety Systems Tested in All Conditions V'Molen NOTE 4 Required by Design Basis Analysis 125.I.6 Valve Torque Limit and Bypass Switch Settings V'Molen NOTE 4 l
125.I.7 Operator Training Adequacy 125.I.7a Recover Failed Equipment V'Molen NOTE 4 125.I.7b Realistic Hands-On Training V'Molen NOTE 4 125.1.8,-
Procedures and Staffing for Reporting to NRC V'Molen NOTE 4 11
l Action lead Lead Office /
Safety Plan Item /
SPEB Division /
Priority Issue No.
Title Engineer Branch Ranking 125.11.1 a Two-Train AlW Reliability V'Molen NOTE 4 125.II. lib Review Existing AFW Systems for Single Failure V'Molen NOTE 4 125.II.1.c NUREG-0737 Reliability Improvements V'Molen NOTE 4 125.II.1.d AFW/ Steam and Feedwater Rupture Control System /ICS V'Molen NOTE 4 Interactions in B&W Plants 125.11.2 Adequacy of Existing Maintenance Requirements for Riggs NOTE 4 Safety-Related Systems 12b.11.3 Review Steam / Feed Line Break V'Molen NOTE 4 125.11.4 OTSG Dryout and Reflood Effects Riggs NOTE 4 125.11.5 Thermal-Hydraulic Effects of loss and Restoration Riggs NOTE 4 of Feedwater on Primary System Components 125.11.6 Reexamine PRA Estimates of Core Damage Risk from V'Molen NOTE 4 1
Steam Generator During Line Break l
125.II.7 Reevaluate Provisions to Automatically Isolate V'Molen NOTE 4 Feedwater from Steam Generator During Line Break
+
l 125.11.8 Reassess Criteria for Feed-and-Bleed Initiation Weinkam NOTE 4 l
125.11.9 Enhance Feed-and-Bleed Capability V'Molen NOTE 4 125.II.1D Hierarchy of Impromptu Operator Actions Riggs NOTE 4 125.11.11 Recovery of Main Feedwater as Alternative to AFW W'einkam NOTE 4 l
125.11.12 Adequacy of Training Regarding PORY Operation Riggs NOTE 4 125.11.14 Remote Operation of Equipment Which Must Now V'Molen NOTE 4 Be Operated Locally 126 Reliability of PWR Main Steam Safety Valves Riggs NOTE 4 127 Testing and Maintenance of Manual Valves in NOTE 4 Safety-Related Systems 1
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t ENCLOSURE 2 TMI ACTION PLAN ITEMS I.A.2 Review operating personnel training to see if accident prevention and mitigation aspects need to be revised and expanded in reference to severe accident sequences.
I.A.4 Consider simulator improvements in terms of simulating plant environmental conditions (e.g., fires, radiation, steam, etc.)
brought about by an accident sequence.
II.A Reconsider the siting of multiple units in view of the potential impact of one unit on another in the event of a severe accident sequence (e.g., limited accessibility of " unaffected" unit due to high radiation fields, physical damage due to explosions, fires, high radiation).
III.A.1 Reexamine the supply and use of thyroid-blocking agents for existing sites.
III.A.3 Consider expansion of emergency preparedness training, planning, and drills, etc., to include multi-state involvement when addressing III.B severe accident sequences. Also, consider increased coordination on the international level.
III.D.1 Some radiation protection areas (e.g., III.D.1.3, III.D.1.3(3),
III.D.1.4) may need to be expanded as more radiation data from the Chernobyl event become available.
III.D.2.5 Consider provisions for including severe accident consequences within the Offsite Dose Calculation Manual (e.g., dose calcu-lation schemes for onsite, site vicinity, continental, and global dispersion).
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UNITED STATES g,
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NUCLEAR REGULATORY COMMISSION
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$31 PARK AVENUE KING OF PRUS$lA. PENNSYLVANIA 1M06
,o 19 JUN 1986 Docket Nos. 50-271 50,-29 Mr. Al Giordano, President MassAlert 52 Grinnell Street Greenfield, MA 01301
Dear Mr. Giordano:
This is to acknowledge receipt of your June 17, 1986 letter to Commissioner Asselstine requesting..an immediate public hearing on the g
operation of the Vermont Yankee Nuclear Station and the Yankee Atomic Plant. A copy of your letter, which' was handed to me during the 'VSNAP meeting in Brattleboro, Vermont, has been forwarded to Commissioner Asselstine and to the NRC Office of Nuclear Reactor Regulation for action.
Sincerely, g,
William F.
ane, Deputy Director Division of Reactor Projects cc:
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6 Commissio er Asselstine A. Bern o, NRR F. Mir glia, NRR
s n.sm o MassAlert Massachusetts Alert
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52 Grinnell Street Greenfield, MA 01301 (413) 772 6098 June 17,198t>
Coasiasioner James Asseletine U.S. Nucl ar Regulatory Conraissi.n Washington, D.C.
20015
Dear Coassissioner,
We are writing to request an insnediate public hearing on the operation of the Vermont Yankee Nucient Station in Vernon, VP and the YarAee Atomic Plant in Rowe. MA.
I As Massachusetts cit"1aene, we live closer to the Vermont plant than most citizens of that state.. The Rowe plant sits within our own Franklin County. Therefore, as citisens or this great nation, we are entitled to such a hearing to have our many questions about these two plants answered.
Serious questions have been raised about the Vermont Yankee plant in both design and the competency of its management. We demand to have thest answered.
The Yankee Atomic Plant in Rowe, as you know is the oldest operating comercial reactor in the United States. You may not be aware that it is the only above ground reactor in the natien. Built in the 1950's before we knew the importance of underground containment, it stands literally as a "Dall on stilts".
The Soviet Nuclear Disaster gave slot of publicity to the need for adequate underground containment. Now, our neighoors have oegun to call the Rowe plant "Chernobyl Yarlee" because it is the only plant in America without such an underground containment.
I will remind you that your own Robert Punie of the U.S. Nuclear Regulatory Commission said on May 20, 1981:
'Being an old plant, if tne present design of Yankee Rowe was drawn up on a piece of paper and autaitted to the N.R.C., it's not likely that they would recieve a license to.da.y".
g This statement is a matter of public record. The transcripts appear in the YarAee Atomic Public Document Room at the Greenfield Consunity. College 1.ibra ry.
We request that this hearing be held not in Vermont - but in Greenfield Massachusetts, our to nty seat. We would be happy to make arrangements for a hearir.g room in the Franklin County Courthouse or some other location.
Copies of this letter are being sent to our frien:!s:
U.S. Rep. Silvio
- 0. Conte ana United States Senators John F. Kerry and Edward M. Kennedy. We request an imediate response and a hearing within the next 30 days.
Sincerely.
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_7 0iordan President, MasaAlert cca Hon. Silvio 0. Conte Hon. Edward M. Kennedy kon. Johti F. Kerry A(^n,emI^tM'.
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