ML20217Q140
| ML20217Q140 | |
| Person / Time | |
|---|---|
| Issue date: | 05/04/1998 |
| From: | Wen P NRC (Affiliation Not Assigned) |
| To: | Essig T NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9805070232 | |
| Download: ML20217Q140 (20) | |
Text
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<> M ou g &
g j UNITED STATES
- NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20565-0001
- + ,o '
May 4, 1998 MEMORANDUM TO: Thomas H. Essig, Acting Chief Generic issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation FROM:
Peter C. Wen, Project Manager f ~ C. I Generic issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation
SUBJECT:
SUMMARY
OF APRIL 23,1998, MEETING WITH THE NUCLEAR ENERGY INSTITUTE (NEI) REGARDING 10 CFR 50.59 On April 23,1998, a public meeting was held at the U.S. Nuclear on's Regulatory Co (NRC's) offices in Rockville, Maryland, between representatives er of the NRC N interested parties. Attachment 1 provides a list of attendees at the meeting A n
includes the agenda that was used for the meeting ande the presentation m y
NEl forofthe discussion meeting.
" safety analysis." Attachments 3 and 4 are supplementalinformation orthe pro three other topics related to the Commission , .
SRM Conceming guidance for sed of March 2 the staff to reconcile their draft guidancegeneric document (NEl 9 letter guidance as soon as possible. The NRR staff members ons stated that such d draft GL (sent to the Commission on April 20,1998).would occur o increases in radiological consequences.notNEl be continues an unreviewed safety question (USQ) if the acceptance limit (such as the P used by the staff to judge acceptability, are still met with the change. The staff ,
it typically performed independent calculations of consequences, rather tha As long as the staff's calculations confirmed ove that the li the facility design and operation. However, the degree of margin might remaining to the be less as viewed by the staff than by the licensee. Therefore, e j if a license changes that would have the effect ofincreasing calculated doses / up to the that the staff conclusion would be that the limits were actually .
a s exceeded NEl s values could be allowed without always requiring g^,~dg NRC 7,
9805070232 980504 PDR REVOP ERGNUMRC PDR O /Y /' " Mf1 -
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T. Essig _ of a recent enforcement action where the change was from 22 Rem to 23 Rem (the limit was 30
~ Rem), as a case that should not have required prior review (and thus which should not have been a violation because it did not).
The next topic was the issue of enforcement discretion. NEl stated their conclusion that enforcement policy changes should be made immediately such that no enforcement action is taken for circumstances that are clearly not safety significant, in order to achieve stability. The staff indicated its plans to continue to exercise discretion with respect to severity levels or issuance of civil penalties (under existing enfo cement policy), pending further interaction with the Commission on enforcement policy changes.
Finally, NEl stated that as part of upcoming rulemaking on 10 CFR 50.59 criteria, the NRC should address the scope of changes that require evaluation directly in the regulation, rather than indirectly through the FSAR. Specifically, they would redefine the changes requiring evaluation against the USQ criteria to be those that affect safety analyses. They would propose to include in the rule a functional definition of " safety analyses" (see preliminary thoughts in Attachment 3), referring to analyses performed pursuant to Commission requirement, or requested to validate conformance with requirements, or other analyses that are approved by -
NRC (by issuance of safety evaluation reports). They would supplement the definition with lists
. of such analyses in a guidance document. A draft outline of how such safety analyses and
^
changes affecting them might be characterized was distributed at the meeting (see Attachment 4). The staff stated that it would consider this proposal but noted that this could not be done on the July 1998 schedule for the proposed rule established by the SRM. Further, the staff emphasized that even if such a rule change were pursued, there is still the need for licensees to update their FSARs to be complete and accurate in accordance with 10 CFR 50.71(e). Further meetings with NEl are anticipated to discuss FSAR update guidance and other issues.
Attachments: As stated cc w/atts: See next page f
- T. Essig . a case that should not have required prior review (and thus which should not have been a violation because it did not).
The next topic was the issue of enforcement discretion. NEl stated their conclusion that enforcement policy changes should be made immediately such that no enforcement action is taken for circumstances that are clearly not safety significant, in order to achieve stability. The staff indicated its plans to continue to exercise discretion with respect to severity levels or issuance of civil penalties (under existing enforcement policy), pending further interaction with the Commission on enforcement policy changes.
Finally, NEl stated that a' s part of upcoming rulemaking on 10 CFR 50.59 criteria, the NRC should address the scope of changes that require evaluation directly in the regulation, rather than indirectly through the FSAR.- Specifically, they would redefene the changes requiring evaluation against the unreviewed safety question (USQ) criteria to ue those that affect safety analyses. They would propose to include in the rule a functional definition of " safety analyses" (see preliminary thoughts in Attachment 3), referring to analyses performed pursuant to Commission requirement, or requested to validate conformance with requirements, or other analyses that are approved by NRC (by issuance ~of safety evaluation reports). They would supplement the definition with lists of such analyses in a guidance document. A draft outline of how such safety analyses and changes affecting them might be characterized was distributed at the meeting (see Attachment 4). The staff stated that it would consider this proposal, but emphasized that even if such a rule change were pursued, there is still the need for licensees to update their FSARs to be complete and accurate in accordance with 10 CFR 50.71(e).
Further meetings with NEl are anticipated to discuss FSAR update guidance and other issues.
Attachments: As statedi _
cc w/atts: See next page ,
CISTRIBUTION: See attached page.
' Document Name: G:\PWX\MSUM.0423.nel - ,
OFFICE: PM:PGEB PGEB a SC:PGEBM7 NAME PWen:swh EMcKeb$N ' FAkstuhE DATE 5/ /98 5// /98 h/98 OFFICAL OFFICE COPY A
_ _-___-----_x_
O Distribution: Mtg. Summary w/ NEl Re 10 CFR 50.59 Scope issue Dated May_4, 1998 Hard_ Copy Dochet Flie PUBLIC PGEB R/F OGC ACRS PWen EMcKenna EMail SCollins/FMiraglia BSheron RZimmerman JRoe DMatthews TEssig FAkstulewicz GMizuro CHolden BHollan THsia KHart GTracey, EDO
NRC/NEl MEETING ON 10 CFR 50.59 ISSUE LIST OF ATTENDEES April 23,1998 NAME ORGANIZAllON David Matthews NRC/NRR/DRPM Tom Essig NRC/NRR/DRPM Frank Akstulewicz NRC/NRR/DRPM Eileen Mckenna NRC/NRR/DRPM Peter Wen NRC/NRR/DRPM Geary Mizuro .- NRC/OGC .
Cornelius Holden NRC/OCM/GID -
Brian Holian NRC/OCM/SAJ Tony Hsia NRC/OCM/NJD Ken Hart NRC/SECY
. Tony Pietrangelo NEl Steve Floyd NEl Doug Walters NEl Russ Bell NEl Nancy Chapman Bechtel Herb Fontecilla . VAP/APS Charlie Brinkman ABB-CE Jerry Dosier NUS Info Services Jenny Weil McGraw Hill Robert Vondrasek PSE&G Sam Crowley Winston & Strawn f Attachment 1
4 NEI Licensing Issues Meeting with NRC April 23,1998
'L*'
Agenda
= FSAR Update Guidance
= Acceptance Limits on Consequences
= Enforcement Discretion related to USQ Determinations
= Scope of 10 CFR 50.59
't* '
Attachment 2 2
9 FSAR Update Guidance !
Objective: Mutually acceptable guidance for utilities ASAP
. Most effective to interact now to
. reconcile industry and NRC draft guidance, per SRM
. then publish . esult (revised NEI 98-03) for public comment
'Y? '
s Status of NEl 98-03
= Distributed for industry comment last November
= No major comments received
= NEI is ready to work with NRC staff now to reconcile with draft GL
'1F '
2
'I s .
Acceptance Limits
= NRC position in Jan. 9 letter to NEI
= Example of the problem
= SRM requests staff to reassess position QEE I Enforcement Discretion
= No enforcement action should be taken during the period prior to the rule change in circumstances that are clearly not safety significant I
= Enforcement policy change should be !
instituted before July 10 i Y'
l 1
3
Purpose of 50.59
= Require licensee review of proposed changes a Determine if change exceeds previously approved design or operational limits l
= Require prior NRC approval if any i authorized limit is exceeded l
'V' v ~..s -
Clarifying the Scope of G 50.59 Principles
= 50.59 isjust one part of a hierarchy ofplant change processes a FSAR is neither appropriate or efficient as the scope of 50 59 4
REGULATORY OVERSIGHT OF PLANT CHANGE CONTROL PROCESS Proposed Change
~
Seek Exemption Meets per 10 CFR 50.12 Regulations or Stop Yes P
Amend License No - '
Per 10 CFR 50.90 hts Operating
- or License? -
Stop Yes P
Seek Amendment to No -
Order per 10 CFR 2.202 Orders?
- or
- Stop Yes Affects Yes Safety Analysis? apply 10 CFR 50.59 No ,,
Change to QA, EP, Process per Security Plan? '
10 CFR 50.54 (a),(p), '
or(q)
No y,, Apply NEl Change to Commitment Commitments?
Management :
Guideline Proceed With Change Update hSAR No ,-
- No Regulatory per 10 CFR 50.71(e)
Interaction Required
i 1
Why change f 50.59(a)(1??
= Too many safety evaluations oflittle or no safety / regulatory value
= Address scope of G 50.59 directly in the regulation, not indirectly via the FSAR
= Improve consistency between rule and i implementation
'F' What are the benefits?
= Clarify the appropriate role and focus of 50.59
= Avoid the need for extensive changes to FSARs, including removal or reformatting ofinformation
= Avoid assigning roles to the FSAR and 50.71(e) for which they are not well suited
. Address concerns about small vs. big FSARs a Facilitate use of acceptance limits criterion for evaluating the effect of changes on consequences
'T' 5
Why now?
= Convergence of 50.59 and FSAR update issues
= Scope issue recognized by industry, NRC staff and Commission
= Include with 50.59 rule changes -- the first in 30 years -- planned for 1998
= More efficient and coherent to address Section a(1) changes in conjunction with other Q 50.59 changes and FSAR update guidance
'17 '
Why Safety Analyses?
= Final exam ofNRC safety review --
principal basis for NRC safety approval
= Provide a nexus to protection of public health and safety
= Encompass design bases
= Only context that makes sense for (a)(2) criteria
't* '
6
i l
How would it work?
= Identify safety analyses l
. from NRC requirements !
. Other analyses approved by SER I
= Identify explicit inputs, assumptions, etc.
= Identify mitigating equipment and operator actions credited i i
= Changes that do not affect analyses would '
screen out >
'1F '
t Summary
= 50.59 enforcement discretion ASAP 1
= Work with NRC staff on
. reconciling draft FSAR update guidance
. 50.59 scope issue !
. reconciling staff comments on NEI 96-07 1
7 l
Proposed Changes to NEI96-07 Include a definition of Safety Analysis SAFETY ANALYSIS A safety analysis is an analysis that is performed pursuant to Commission requirements or requested by NRC to .
t validate compliance with existing requirements, and is necessary to demonstrate the integrity of the reactor coolant pressure boundary, the capability to shutdown the reactor and maintain it in a safe shutdown condition, and the capability to prevent or mitigate accidents that could result in potential offsite exposures.
Safety an dvses include:
l analyses included in the FSAR and approved by the Commission as part ofinitial i licensing :
analyses performed pursuant to new or amended Commission regulations l
subsequent to initial licensing '
analyses performed in response to a generic or plant-specific issue to validate compliaace with existing requirements analyses specifically approved by the NRC via SER Note: When a new analysis or change to plant orprocedures "affects"one or more safety analyses, the safety analyses should be updated to reflect the change to maintain an accurate baseline for evaluation offuture changes.
Safety analyses do not include:
detailed calculations and other non-docketed analyses performed in support of safety analyses environmental, financial and other analyses unrelated to nuclear safety docketed information controlled by other regulations (QA, EP, Security) analyses submitted to the NRC in response to generic communications that do not affect analyses required to support initial licensing or demonstrate compliance with new or amended regulations (Note: required analyses should be updated to reflect the effects of other changes, analyses or issues.)
analyses provided in LER or NOV responses except as required to demonstrate compliance with NRC regulations; the effects of such analyses should be incorporated in the UFSAR in a subsequent update Attachment 3 l
l DRAFT ,
Identification of Safety Analyses ;
i Safety Analysis Basis for NRC Safety Analysis SER or other NRC Approval !
Requirement ' Reference
- 1. General GDC I
- 4. Increase in Steam Flow GDC 10,15,26
- 5. Inadvertent Steam Generator Safety or GDC 10,15,26 Relief Valve Opening (PWR) J
- 6. Steam System Piping Failure inside and GDC 27,28,31, Outside of Containment (PWR) 35,10 CFR 100
- 7. Loss of External Load GDC 10,15,26 l
- 8. Turbine Trip GDC 10,15,26
- 9. Loss of Condenser Vacuum GDC 10,15,26
- 12. FW System Pipe Breaks inside and GDC 27,28,31 Outside Containment (PWR) 35,10 CFR 100 1
- 13. Loss of Coolant Flow Including Pump GDC 10,15,26
.l Trip i
- 14. Reactor Coolant Pump Rotor Seizure GDC 27,28,31, j 10 CFR 100 j
- 15. Reactor Coolant Pump Shaft Break GDC 27,28,31, 10 CFR 100
- 16. Uncontrolled Rod Withdrawal from a GDC 10,20,25 Suberitical or Low Power Condition
- 17. Uncontrolled Rod Withdrawal at Power GDC 10,20,25
- 18. Control Rod Misoperation (System GDC 10,20,25 Malfunction or Operator Error)
- 19. Startup of an inactive or Recirculation GDC 10,15,20, I Loop at an incorrect Temperature 26,28 )
- 20. CVCS Malfunction that Results in a GDC 10,15,26 Decrease in the Boron Concentration in the Reactor Coolant (PWR)
- 21. Inadvertent Loading and Operation of a GDC 13, Fuel Assembly in a Improper Position 10 CFR 100
- 22. Spectrum of Rod Ejection Accidents GDC 28, (PWR) 10 CFR 100
- 24. CVCS Malfunction that increases GDC 10,15,26 Reactor Coolant inventory (PWR)
- 26. Radiological Consequences of the GDC55, Failure of Small Lines Carrying PWR 10 CFR 100 Primary Coolant Outside Containment
- 27. Radiological Consequences of a Steam 10 CFR 100 Generator Tube Failure (PWR)
Attachment 4
l.
- 28. LOCAs Resulting from Spectrum of 10 CFR 50.46,
, Postulated Piping Breaks within the App. K, GDC 35, j Reactor Coolant Pressure Boundary 10 CFR 100 l 29. Radioactive Liquid Waste System Leak or Failure (Release to the Atmosphere)
- 30. l Radioactive Gas Waste System Leak or Failure
- 32. Radiological Consequences of Fuel GDC 61, Handling Accidents 10 CFR 100
- 33. Spent Fuel Cask Drop Accidents GDC 61, 10 CFR 100
- 34. Containment Analysis GDC 50
- 35. Power Uprate Analysis NA
- 38. MOV CMFs during Transients due to IEB 85-03 Improper Switch Settings
- 41. Resolution of GI A 30, Adequacy of S. GL 91-06 R DC Power Supplies,10 CFR 50.54(O
- 42. Reactor Vessel Structural Integrity GL 92-01
- 43. WEC Rod Control System Failure and GL 93-04 Withdrawal of RCCAs,10 CFR 50.54(O
- 46. Anticipated Transients Without Scram 10 CFR 50.62, GDC 10,15,26, 27,29
- 47. Pressurized Thermal Shock 10 CFR 50.61
- 48. Station Blackout 10 CFR 50.63 I- 49. Fire Protection Appendix R
- 50. Environmental Qualification 10 CFR 50.49
- 51. TMI Items 10CFR 50.34(f) l l
i DRAFT Generic Conununications That May Have Led To New Analyses Bulletins Bulletin
- Title Comment
- 2. IEB 96-02 Movement of Heavy Loads Over Spent Fuel, Over Fuelin the Reactor ALL Core, or Over Safety Related Equipment
- 3. IEB 96 01 Control Rod Insertion Problems WESTINGHOUSE
- 6. IEB 89-03 Potential less Of Required Shutdown Margin During Refueling PWR Operations
- 7. IEB 8811 Fressurizer Surge Line Thermal Stratification PWR
- 8. IEB 88-08 Tisermal Stresses In Piping Connected To Reactor Coolant Systems ALL
- 10. IEB 88-04 Potential Safety Related Pump less ALL
- 11. IEB 88-02 Rapidly Proparating Fatigue Cracks In Steam Generator Tubes WESTINGHOUSE
- 12. IEB 85 03 Motor-Operated Valve Common Mode Failures During Plant ALL ,
Transients Due To Improper Switch Settings i
13 IEB 84 03 Refueling Cavity Water Seal ALL 1
- 14. IEB 83-07 Apparently Fraudulent Products Sold By Ray Miller, Inc. ALL
- 15. IEB 8102 Fadure Of Gate Type Valves To Close Against Differential Pressure ALL
- 16. IEB 80 23 Fadures Of Solenoid Valves Manufactured By Valcor Engineering ALL I Convoration
]
- 17. IEB 80-18 Matatenance Of Adequate Minimum Flow Through Centrifugal PWR Charging Pumps Following Secondary Side High Enerry Line Rupture
- 18. IEB 80-17 Fadure Of Control Rods To Insert During A Scram At A BWR BWR
- 19. IEB 80-16 Potential Misapplication Of Rosemount Inc., Models 1151 And 1152 ALL Pressure Transmitters With Either "A" Or "B" Output Codes
- 20. IEB 80-11 Masonry Wall Design ALL
- 22. IEB 80 04 Analysis Of A PWR Main Steam Line Break With Continued PWR Teedwater Addition
- 23. IEB 79 27 Loss Of Non Class IE Instrumentation And Control Power Systems ALL Bus During Operation
- 25. IEB 7914 Seismic Analysis For As-Built Safety Related Piping Systems PWR
- 27. IEB 79-07 Seismic Stress Analysis Of Safety Related Piping ALL
- 28. IEB 79 02 Pipe Support Base Plate Designs Using Concrete Expansion Anchor ALL Bolts
- 29. IEB 79 01 Environmental QualiSeation Of Class le Equipment ALL l
l i
I'
! DRAFT l
Generic Conununications That May Have Led To New Analyses Generic Letters Generic Lir Title Comment j 1. GL 97 04 Assurance Of Sufficient Net Positive Suction Head For Emergency ALL Core Cooling And Containment Heat Removal Pumps l 2. GL 96-06 Assurance Of Equipment Operability And Containment Integnty ALL During Design Basis Accident Conditions
- 3. GL 96-04 BoraDex Degradation In Spent Fuel Pool Storage Racks ALL
- 4. GL 95-07 Pressure Locking And Thermal Binding Of Safety. Related Power. ALL Operated Gate Valves
- 5. GL 95-03 Circumferential Cracking Of Steam Generator Tubes PWR
- 6. GL 94-03 Intergranular Stress Corrosion Cracking Of Core Shrouds In Boiling BWR l Water Reactors
- 7. GL 93-04 Rod Control System Failure And Withdrawal Of Rod Control Cluster WESTINGHOUSE
, Assemblies,10 CFR 50.54(F) l 8. GL 92 04 Resolution Of The Issues Related To Reactor Vessel Water Level BWR Instrumentation In BWRs Pursuant To 10 CFR 50.54(F)
- 9. GL 92 01 Reactor Vessel Structural Integrity ALL
- 10. GL 91-06 Resolution Of Generic Issue A 30, " Adequacy Of Safety-Related DC ALL (No direet Power Supplies." Pursuant To 10 CFR 50.54(F) response required) i 11. GL 89 21 Request For Information Concerning Status OfImplementation Of ALL Unresolved Safety Issue (USI) Requirements
- 12. GL 89-10 Safety Related (1) Motor-Operated Valve Testing And Surveillance ALL l 13. GL 88-20 Individual Plant Examination Of External Events For Severe Accident ALL Vulnerabilities
- 14. GL 8814 Instrument Air Supply System Problems Affecting Safety Related ALL Equipment
- 15. GL 88-01 NRC Position On IGSCC In BWR Austenitic Stainless Steel Piping BWR l 16. GL 8132 Nureg 0737, Item li.K.3.44 Evaluation Of Anticipated Transients BWR (Referenemg Combined With Single Failure BWROG response to NUREG 0737 II k.3.44
- 18. GL 81 14 Seismic Qualification Of Auxiliary Feedwater Systems PWR
- 19. GL 81 12 Fire Protection Rule (45 F/R 76602, November 19,1980) ALL (Licensed prior to 1/1/79)
- 20. GL 8107 Control Of Heavy leads ALL
- 21. GL 78 09 Multiple Subsequent Actuations Of Safety / Relief Valves Following An BWR Isolation Event l
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l Nuclear Energy Institute . Project No. 689
. cc: Mr. Ralph Beedle Ms. Lynnette Hendricks, Director Senior Vice President Plant Support and Chief Nuclear Omcer - Nuclear Energy Institute Nuclear Energy Institute _ Suite 400 -
Suite 400 1776 l Street, NW 1776 i Street, NW Washington, DC 20006-3708 Washington, DC 20006-3708
{
j Mr. Alex Marion, Director l Programs' ,
Nuclear Energy Institute Suite 400 1776 i Street, NW .
Washington, DC 20006-3708 Mr. David Modeen, Director Engineering Nuclear Energy Institute Suite 400 1776 i Street, NW i Washington, DC 20006-3708 Mr. Anthony Pietrangelo, Director Licensing
. Nuclear Energy institute L
Suite 400 1776 l Street, NW -
Washington, DC 20006-3708 Mr. Nicholas J. Liparulo, Manager Nuclear Safety and Regulatory Activities Nuclear and Advanced Technology Division Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, Pennsylvania 15230
' Mr. Jim Davis, Director Operations Nuclear Energy Institute Suite 400 -
- 1776 l Street, NW .
Washington, DC 20006-3708:
t