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=Text=
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            <> M ou g             &
<> M ou g
g                j                                  UNITED STATES
UNITED STATES g
                    #                  NUCLEAR       REGULATORY COMMISSION WASHINGTON, D.C. 20565-0001
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20565-0001
          ****+ ,o                                                                                                                 '
****+,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:
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
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:==
==SUBJECT:==
==SUMMARY==
==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
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 Regulatory Co (NRC's) offices in Rockville, Maryland, between representatives of the NRC N on's interested parties. Attachment 1 provides a list of attendees at the meeting A er includes the agenda that was used for the meeting and the presentation m n
includes the agenda that was used for the meeting ande the                                                   presentation m y
NEl for the meeting. Attachments 3 and 4 are supplementalinformation pro e
NEl forofthe discussion            meeting.
y discussion of " safety analysis."
                      " safety   analysis." Attachments 3 and 4 are supplementalinformation                orthe pro three other topics related to the Commission                           ,      .
orthe three other topics related to the Commission SRM of March 2 sed the staff to reconcile their draft guidance document (NEl Conceming guidance for letter guidance as soon as possible. The NRR staff members stated that such d generic draft GL (sent to the Commission on April 20,1998).would occur o ons increases in radiological consequences. NEl continues an unreviewed safety question (USQ) if the acceptance limit (such as the P not be 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 that the li the facility design and operation. However, the degree of margin remaining to the ove be less as viewed by the staff than by the licensee. Therefore, if a license might changes that would have the effect ofincreasing calculated doses up to the j
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                         ,
e that the staff conclusion would be that the limits were actually exceeded NEl s
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,
values could be allowed without always requiring NRC g^,~dg a
9805070232 980504 PDR       REVOP ERGNUMRC PDR                                    O /Y /' "         Mf1   -
s 7,
g            j j'>yQ g
9805070232 980504
,d4      '
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PDR REVOP ERGNUMRC O Y /' "
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,d4 j j'>yQ g g
ya
ya


T. Essig _                                                   of a recent enforcement action where the change was from 22 Rem to 23 Rem (the limit was 30
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).
~ 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.
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 -
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 ), 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
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
. 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.
changes affecting them might be characterized was distributed at the meeting (see ). 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
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).
- 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.
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).
Finally, NEl stated that ' s part of upcoming rulemaking on 10 CFR 50.59 criteria, the NRC a
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.
Further meetings with NEl are anticipated to discuss FSAR update guidance and other issues.
Attachments: As statedi                                         _
Attachments: As statedi cc w/atts: See next page CISTRIBUTION: See attached page.
cc w/atts: See next page             ,
' Document Name: G:\\PWX\\MSUM.0423.nel -
CISTRIBUTION: See attached page.
SC:PGEBM7 OFFICE:
        ' Document Name: G:\PWX\MSUM.0423.nel -                             ,
PM:PGEB PGEB a PWen:swh EMcKeb$N '
OFFICE:                 PM:PGEB       PGEB   a       SC:PGEBM7 NAME                    PWen:swh EMcKeb$N ' FAkstuhE DATE                     5/ /98         5// /98       h/98 OFFICAL OFFICE COPY A
FAkstuhE NAME DATE 5/ /98 5// /98 h/98 OFFICAL OFFICE COPY A
_    _-___-----_x_
_-___-----_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
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 .
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.-
Cornelius Holden                 NRC/OCM/GID -
NRC/OGC.
Brian Holian                     NRC/OCM/SAJ Tony Hsia                       NRC/OCM/NJD Ken Hart                         NRC/SECY
Cornelius Holden NRC/OCM/GID -
    . 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
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


4 NEI Licensing Issues Meeting with NRC April 23,1998
4 NEI Licensing Issues Meeting with NRC April 23,1998
                                                      'L*'
'L*'
Agenda
Agenda
          = FSAR Update Guidance
= FSAR Update Guidance Acceptance Limits on Consequences
          = Acceptance Limits on Consequences
=
          = Enforcement Discretion related to USQ Determinations
Enforcement Discretion related to
          = Scope of 10 CFR 50.59
=
                                                      't* '
USQ Determinations Scope of 10 CFR 50.59
Attachment 2 2
=
't* '
2


9 FSAR Update Guidance                                             !
9 FSAR Update Guidance Objective: Mutually acceptable guidance for utilities ASAP
Objective: Mutually acceptable guidance for utilities ASAP
. Most effective to interact now to
          . Most effective to interact now to
. reconcile industry and NRC draft guidance, per SRM
              . reconcile industry and NRC draft guidance, per SRM
. then publish. esult (revised NEI 98-03) for public comment
              . then publish . esult (revised NEI 98-03) for public comment
'Y? '
                                                            'Y? '
s Status of NEl 98-03 Distributed for industry comment last
s Status of NEl 98-03
=
        = Distributed for industry comment last November
November No major comments received
        = No major comments received
=
        = NEI is ready to work with NRC staff now to reconcile with draft GL
NEI is ready to work with NRC staff
                                                          '1F '
=
now to reconcile with draft GL
'1F '
2
2


                'I s     .
'I s
Acceptance Limits
Acceptance Limits
      = NRC position in Jan. 9 letter to NEI
= NRC position in Jan. 9 letter to NEI
      = Example of the problem
= Example of the problem
      = SRM requests staff to reassess position QEE I Enforcement Discretion
= 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
= No enforcement action should be taken during the period prior to the rule change in circumstances that are clearly not safety significant
      = Enforcement policy change should be           !
= Enforcement policy change should be instituted before July 10 i
instituted before July 10                     i Y'
Y' 3
l 1
3


Purpose of 50.59
Purpose of 50.59
    = Require licensee review of proposed changes a Determine if change exceeds previously approved design or operational limits l
= Require licensee review of proposed changes a Determine if change exceeds previously approved design or operational limits
    = Require prior NRC approval if any             i authorized limit is exceeded                   l
= Require prior NRC approval if any i
                                              'V' v ~..s                       -
authorized limit is exceeded
Clarifying the Scope of G 50.59 Principles
'V' v ~..s Clarifying the Scope of G 50.59 Principles 50.59 isjust one part of a hierarchy
    =  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
=
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
REGULATORY OVERSIGHT OF PLANT CHANGE CONTROL PROCESS Proposed Change Seek Exemption
                                          ~
~
Seek Exemption Meets                                   per 10 CFR 50.12 Regulations                     or Stop Yes P
Meets per 10 CFR 50.12 Regulations or Stop Yes P
Amend License No       -                                                                                                                                                  '
Amend License No Per 10 CFR 50.90 hts or Operating License?
Per 10 CFR 50.90 hts Operating
:  or License?                           -
Stop Yes P
Stop Yes P
Seek Amendment to No             -
Seek Amendment to No Order per 10 CFR 2.202 or Orders?
Order per 10 CFR 2.202 Orders?
Stop Yes Affects Yes Safety Analysis?
:  or
apply 10 CFR 50.59 No Change to QA, EP, Process per Security Plan?
:    Stop Yes Affects           Yes Safety Analysis?                             apply 10 CFR 50.59 No       ,,
10 CFR 50.54 (a),(p),
Change to QA, EP,                           Process per Security Plan?               '
10 CFR 50.54 (a),(p),                                                                                                                           '
or(q)
or(q)
No y,,                 Apply NEl Change to                           Commitment Commitments?
No Apply NEl y,,
Management                                                                                                                                 :
Change to Commitment Commitments?
Guideline Proceed With Change                                                                                                                     Update hSAR No   ,-
Management Guideline Proceed With Change Update hSAR No No Regulatory per 10 CFR 50.71(e)
:        No Regulatory                                                                                                                     per 10 CFR 50.71(e)
Interaction Required
Interaction Required


i 1
i Why change f 50.59(a)(1??
Why change f 50.59(a)(1??
= Too many safety evaluations oflittle or no safety / regulatory value
    = 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
    = Address scope of G 50.59 directly in the regulation, not indirectly via the FSAR
= Improve consistency between rule and i
    = Improve consistency between rule and                 i implementation
implementation
                                                      'F' What are the benefits?
'F' What are the benefits?
    = Clarify the appropriate role and focus of 50.59
= Clarify the appropriate role and focus of 50.59
    = Avoid the need for extensive changes to FSARs, including removal or reformatting ofinformation
= 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
= 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
. Address concerns about small vs. big FSARs a Facilitate use of acceptance limits criterion for evaluating the effect of changes on consequences
                                                      'T' 5
'T' 5


Why now?
Why now?
  = Convergence of 50.59 and FSAR update issues
= Convergence of 50.59 and FSAR update issues
  = Scope issue recognized by industry, NRC staff and Commission
= Scope issue recognized by industry, NRC staff and Commission
  = Include with 50.59 rule changes -- the first in 30 years -- planned for 1998
= 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
= More efficient and coherent to address Section a(1) changes in conjunction with other Q 50.59 changes and FSAR update guidance
                                                '17 '
'17 '
Why Safety Analyses?
Why Safety Analyses?
  = Final exam ofNRC safety review --
= Final exam ofNRC safety review --
principal basis for NRC safety approval
principal basis for NRC safety approval
  = Provide a nexus to protection of public health and safety
= Provide a nexus to protection of public health and safety
  = Encompass design bases
= Encompass design bases
  = Only context that makes sense for (a)(2) criteria
= Only context that makes sense for (a)(2) criteria
                                                't* '
't* '
6
6


i l
i How would it work?
How would it work?
= Identify safety analyses
      = Identify safety analyses                           l
. from NRC requirements
        . from NRC requirements                           !
. Other analyses approved by SER I
        . Other analyses approved by SER I
= Identify explicit inputs, assumptions, etc.
      = Identify explicit inputs, assumptions, etc.
= Identify mitigating equipment and operator actions credited i
      = Identify mitigating equipment and operator actions credited                                   i i
i
      = Changes that do not affect analyses would           '
= Changes that do not affect analyses would screen out
screen out                                         >
'1F '
                                                  '1F '
t Summary 50.59 enforcement discretion ASAP
t Summary
=
      =    50.59 enforcement discretion ASAP 1
1
      = Work with NRC staff on
= Work with NRC staff on
        . reconciling draft FSAR update guidance
. reconciling draft FSAR update guidance 50.59 scope issue
        . 50.59 scope issue                           !
. reconciling staff comments on NEI 96-07 1
        . reconciling staff comments on NEI 96-07 1
7 l
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           .
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
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.
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:
Safety an dvses include:
l analyses included in the FSAR and approved by the Commission as part ofinitial         i licensing                                                                           :
analyses included in the FSAR and approved by the Commission as part ofinitial licensing analyses performed pursuant to new or amended Commission regulations 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.
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:
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.)
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
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 l


l DRAFT                                                       ,
DRAFT Identification of Safety Analyses i
Identification of Safety Analyses                                         ;
Safety Analysis Basis for NRC Safety Analysis SER or other NRC Approval Requirement '
i Safety Analysis             Basis for NRC Safety Analysis SER or other NRC Approval     !
Reference 1.
Requirement '   Reference
General GDC 2.
: 1. General                                 GDC                                                           I
Decrease in FW Temperature GDC 10,15,26 3.
: 2. Decrease in FW Temperature             GDC 10,15,26                                                   !
Increase in FW Flow GDC 10,15,26 4.
: 3. Increase in FW Flow                     GDC 10,15,26
Increase in Steam Flow GDC 10,15,26 5.
: 4. Increase in Steam Flow                 GDC 10,15,26
Inadvertent Steam Generator Safety or GDC 10,15,26 Relief Valve Opening (PWR)
: 5. Inadvertent Steam Generator Safety or   GDC 10,15,26 Relief Valve Opening (PWR)                                                                             J
J 6.
: 6. Steam System Piping Failure inside and GDC 27,28,31, Outside of Containment (PWR)           35,10 CFR 100
Steam System Piping Failure inside and GDC 27,28,31, Outside of Containment (PWR) 35,10 CFR 100 7.
: 7. Loss of External Load                   GDC 10,15,26                                                   l
Loss of External Load GDC 10,15,26 l
: 8. Turbine Trip                           GDC 10,15,26
8.
: 9. Loss of Condenser Vacuum               GDC 10,15,26
Turbine Trip GDC 10,15,26 9.
: 10. Loss of Non-emergency AC Power to       GDC 10,15,26 the Station Auxiliaries                                                                               i
Loss of Condenser Vacuum GDC 10,15,26 10.
: 11. Loss of Normal FW Flow                 GDC 10,15,26
Loss of Non-emergency AC Power to GDC 10,15,26 the Station Auxiliaries i
: 12. FW System Pipe Breaks inside and       GDC 27,28,31 Outside Containment (PWR)               35,10 CFR 100                                                 1
11.
: 13. Loss of Coolant Flow Including Pump     GDC 10,15,26
Loss of Normal FW Flow GDC 10,15,26 12.
                                                                                                                .l Trip                                                                                                   i
FW System Pipe Breaks inside and GDC 27,28,31 Outside Containment (PWR) 35,10 CFR 100 13.
: 14. Reactor Coolant Pump Rotor Seizure     GDC 27,28,31,                                                 j 10 CFR 100                                                     j
Loss of Coolant Flow Including Pump GDC 10,15,26
: 15. Reactor Coolant Pump Shaft Break       GDC 27,28,31, 10 CFR 100
.l Trip i
: 16. Uncontrolled Rod Withdrawal from a     GDC 10,20,25 Suberitical or Low Power Condition
14.
: 17. Uncontrolled Rod Withdrawal at Power   GDC 10,20,25
Reactor Coolant Pump Rotor Seizure GDC 27,28,31, j
: 18. Control Rod Misoperation (System       GDC 10,20,25 Malfunction or Operator Error)
10 CFR 100 j
: 19. Startup of an inactive or Recirculation GDC 10,15,20,                                                 I Loop at an incorrect Temperature       26,28                                                         )
15.
: 20. CVCS Malfunction that Results in a     GDC 10,15,26 Decrease in the Boron Concentration in the Reactor Coolant (PWR)
Reactor Coolant Pump Shaft Break GDC 27,28,31, 10 CFR 100 16.
: 21. Inadvertent Loading and Operation of a GDC 13, Fuel Assembly in a Improper Position   10 CFR 100
Uncontrolled Rod Withdrawal from a GDC 10,20,25 Suberitical or Low Power Condition 17.
: 22. Spectrum of Rod Ejection Accidents     GDC 28, (PWR)                                   10 CFR 100
Uncontrolled Rod Withdrawal at Power GDC 10,20,25 18.
;      23. Inadvertent Operation of ECCS           GDC 10,15,26
Control Rod Misoperation (System GDC 10,20,25 Malfunction or Operator Error) 19.
: 24. CVCS Malfunction that increases         GDC 10,15,26 Reactor Coolant inventory (PWR)
Startup of an inactive or Recirculation GDC 10,15,20, Loop at an incorrect Temperature 26,28
: 25. Inadvertent Opening of a FWR Pr.       GDC 10,15,26 Relief Valve or a BWR Relief Valve
)
: 26. Radiological Consequences of the       GDC55, Failure of Small Lines Carrying PWR     10 CFR 100 Primary Coolant Outside Containment
20.
: 27. Radiological Consequences of a Steam   10 CFR 100 Generator Tube Failure (PWR)
CVCS Malfunction that Results in a GDC 10,15,26 Decrease in the Boron Concentration in the Reactor Coolant (PWR) 21.
Attachment 4
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 23.
Inadvertent Operation of ECCS GDC 10,15,26 24.
CVCS Malfunction that increases GDC 10,15,26 Reactor Coolant inventory (PWR) 25.
Inadvertent Opening of a FWR Pr.
GDC 10,15,26 Relief Valve or a BWR Relief Valve 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)  


l.
l.
: 28. LOCAs Resulting from Spectrum of       10 CFR 50.46,
28.
,        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)
LOCAs Resulting from Spectrum of 10 CFR 50.46, Postulated Piping Breaks within the App. K, GDC 35, j
: 30.                                                         l Radioactive Gas Waste System Leak or Failure
Reactor Coolant Pressure Boundary 10 CFR 100 l
: 31. Postulated Radioactive Release due to GDC 60, Liquid-Containing Tank Failures         10 CFR 20
29.
: 32. Radiological Consequences of Fuel     GDC 61, Handling Accidents                     10 CFR 100
Radioactive Liquid Waste System Leak or Failure (Release to the Atmosphere) 30.
: 33. Spent Fuel Cask Drop Accidents         GDC 61, 10 CFR 100
Radioactive Gas Waste System Leak or Failure 31.
: 34. Containment Analysis                   GDC 50
Postulated Radioactive Release due to GDC 60, Liquid-Containing Tank Failures 10 CFR 20 32.
: 35. Power Uprate Analysis                   NA
Radiological Consequences of Fuel GDC 61, Handling Accidents 10 CFR 100 33.
: 36. Temperature Effects on PWR Level       IEB 79-21 Measurements
Spent Fuel Cask Drop Accidents GDC 61, 10 CFR 100 34.
: 37. Analysis of a PWR MSL Break with       IEB 80-04       j Continued Feedwater Addition                           i
Containment Analysis GDC 50 35.
: 38. MOV CMFs during Transients due to       IEB 85-03 Improper Switch Settings
Power Uprate Analysis NA 36.
: 39. Pressurizer Surge Line Thermal         IEB 88-11 Stratification in PWRs
Temperature Effects on PWR Level IEB 79-21 Measurements 37.
: 40. Seismic Qualification Of Auxiliary     GL 81-14 Feedwater Systems l
Analysis of a PWR MSL Break with IEB 80-04 j
: 41. Resolution of GI A 30, Adequacy of S. GL 91-06 R DC Power Supplies,10 CFR 50.54(O
Continued Feedwater Addition i
: 42. Reactor Vessel Structural Integrity     GL 92-01
38.
: 43. WEC Rod Control System Failure and     GL 93-04 Withdrawal of RCCAs,10 CFR 50.54(O
MOV CMFs during Transients due to IEB 85-03 Improper Switch Settings 39.
: 44. Equipment Operability / Containment     GL 96-06 Integrity under DBA Conditions
Pressurizer Surge Line Thermal IEB 88-11 Stratification in PWRs 40.
: 45. Assurance of Sufficient NPSH for ECC   GL 97-04 and Containment Heat Removal Pumps
Seismic Qualification Of Auxiliary GL 81-14 Feedwater Systems l
: 46. Anticipated Transients Without Scram   10 CFR 50.62, GDC 10,15,26, 27,29
41.
: 47. Pressurized Thermal Shock               10 CFR 50.61
Resolution of GI A 30, Adequacy of S.
: 48. Station Blackout                       10 CFR 50.63 I-   49. Fire Protection                         Appendix R
GL 91-06 R DC Power Supplies,10 CFR 50.54(O 42.
: 50. Environmental Qualification             10 CFR 50.49
Reactor Vessel Structural Integrity GL 92-01 43.
: 51. TMI Items                               10CFR 50.34(f) l l
WEC Rod Control System Failure and GL 93-04 Withdrawal of RCCAs,10 CFR 50.54(O 44.
Equipment Operability / Containment GL 96-06 Integrity under DBA Conditions 45.
Assurance of Sufficient NPSH for ECC GL 97-04 and Containment Heat Removal Pumps 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
i DRAFT Generic Conununications That May Have Led To New Analyses Bulletins Bulletin
* Title                                                             Comment
* Title Comment 1.
: 1. IEB 96-03     , Potential Plugging of Emergency Core Cooling Suction Strainers by BWR Debris in BWRs
IEB 96-03
: 2. IEB 96-02       Movement of Heavy Loads Over Spent Fuel, Over Fuelin the Reactor ALL Core, or Over Safety Related Equipment
, Potential Plugging of Emergency Core Cooling Suction Strainers by BWR Debris in BWRs 2.
: 3. IEB 96 01       Control Rod Insertion Problems                                   WESTINGHOUSE
IEB 96-02 Movement of Heavy Loads Over Spent Fuel, Over Fuelin the Reactor ALL Core, or Over Safety Related Equipment 3.
: 4. IEB 93-02       Debris Plurring of ECCS Suction Strainers                         ALL
IEB 96 01 Control Rod Insertion Problems WESTINGHOUSE 4.
: 5. IEB 90-02       Lose Of Thermal Marrin Caused By Channel Box Bow                 BWR
IEB 93-02 Debris Plurring of ECCS Suction Strainers ALL 5.
: 6. IEB 89-03       Potential less Of Required Shutdown Margin During Refueling       PWR Operations
IEB 90-02 Lose Of Thermal Marrin Caused By Channel Box Bow BWR 6.
: 7. IEB 8811       Fressurizer Surge Line Thermal Stratification                     PWR
IEB 89-03 Potential less Of Required Shutdown Margin During Refueling PWR Operations 7.
: 8. IEB 88-08       Tisermal Stresses In Piping Connected To Reactor Coolant Systems ALL
IEB 8811 Fressurizer Surge Line Thermal Stratification PWR 8.
: 9. IEB 88-07       Power Oscillations In Boiling Water Reactors (BWR)               BWR
IEB 88-08 Tisermal Stresses In Piping Connected To Reactor Coolant Systems ALL 9.
: 10. IEB 88-04       Potential Safety Related Pump less                               ALL
IEB 88-07 Power Oscillations In Boiling Water Reactors (BWR)
: 11. IEB 88-02       Rapidly Proparating Fatigue Cracks In Steam Generator Tubes       WESTINGHOUSE
BWR 10.
: 12. IEB 85 03       Motor-Operated Valve Common Mode Failures During Plant           ALL         ,
IEB 88-04 Potential Safety Related Pump less ALL 11.
Transients Due To Improper Switch Settings i
IEB 88-02 Rapidly Proparating Fatigue Cracks In Steam Generator Tubes WESTINGHOUSE 12.
13 IEB 84 03       Refueling Cavity Water Seal                                       ALL         1
IEB 85 03 Motor-Operated Valve Common Mode Failures During Plant ALL Transients Due To Improper Switch Settings i
: 14. IEB 83-07       Apparently Fraudulent Products Sold By Ray Miller, Inc.           ALL
13 IEB 84 03 Refueling Cavity Water Seal ALL 1
: 15. IEB 8102       Fadure Of Gate Type Valves To Close Against Differential Pressure ALL
14.
: 16. IEB 80 23       Fadures Of Solenoid Valves Manufactured By Valcor Engineering     ALL         I Convoration
IEB 83-07 Apparently Fraudulent Products Sold By Ray Miller, Inc.
                                                                                                        ]
ALL 15.
: 17. IEB 80-18       Matatenance Of Adequate Minimum Flow Through Centrifugal         PWR Charging Pumps Following Secondary Side High Enerry Line Rupture
IEB 8102 Fadure Of Gate Type Valves To Close Against Differential Pressure ALL 16.
: 18. IEB 80-17       Fadure Of Control Rods To Insert During A Scram At A BWR         BWR
IEB 80 23 Fadures Of Solenoid Valves Manufactured By Valcor Engineering ALL Convoration
: 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
17.
: 21. IEB 80-07       BWR Jet Pump Assembly Failure                                     BWR
IEB 80-18 Matatenance Of Adequate Minimum Flow Through Centrifugal PWR Charging Pumps Following Secondary Side High Enerry Line Rupture 18.
: 22. IEB 80 04       Analysis Of A PWR Main Steam Line Break With Continued           PWR Teedwater Addition
IEB 80-17 Fadure Of Control Rods To Insert During A Scram At A BWR BWR 19.
: 23. IEB 79 27       Loss Of Non Class IE Instrumentation And Control Power Systems   ALL Bus During Operation
IEB 80-16 Potential Misapplication Of Rosemount Inc., Models 1151 And 1152 ALL Pressure Transmitters With Either "A" Or "B" Output Codes 20.
: 24. IEB 79-21   ,   Temperature Effects On Level Measurements                         PWR
IEB 80-11 Masonry Wall Design ALL 21.
: 25. IEB 7914       Seismic Analysis For As-Built Safety Related Piping Systems       PWR
IEB 80-07 BWR Jet Pump Assembly Failure BWR 22.
: 26. IEB 79-12       Short Period Scrams At BWR Facilities                             BWR
IEB 80 04 Analysis Of A PWR Main Steam Line Break With Continued PWR Teedwater Addition 23.
: 27. IEB 79-07       Seismic Stress Analysis Of Safety Related Piping                 ALL
IEB 79 27 Loss Of Non Class IE Instrumentation And Control Power Systems ALL Bus During Operation 24.
: 28. IEB 79 02       Pipe Support Base Plate Designs Using Concrete Expansion Anchor   ALL Bolts
IEB 79-21
: 29. IEB 79 01       Environmental QualiSeation Of Class le Equipment                 ALL l
, Temperature Effects On Level Measurements PWR 25.
IEB 7914 Seismic Analysis For As-Built Safety Related Piping Systems PWR 26.
IEB 79-12 Short Period Scrams At BWR Facilities BWR 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
l i


I'
I' DRAFT l
!                                                      DRAFT l
Generic Conununications That May Have Led To New Analyses Generic Letters Generic Lir Title Comment j
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
1.
: 3. GL 96-04   BoraDex Degradation In Spent Fuel Pool Storage Racks                 ALL
GL 97 04 Assurance Of Sufficient Net Positive Suction Head For Emergency ALL Core Cooling And Containment Heat Removal Pumps l
: 4. GL 95-07   Pressure Locking And Thermal Binding Of Safety. Related Power.       ALL Operated Gate Valves
2.
: 5. GL 95-03   Circumferential Cracking Of Steam Generator Tubes                     PWR
GL 96-06 Assurance Of Equipment Operability And Containment Integnty ALL During Design Basis Accident Conditions 3.
: 6. GL 94-03   Intergranular Stress Corrosion Cracking Of Core Shrouds In Boiling   BWR l                             Water Reactors
GL 96-04 BoraDex Degradation In Spent Fuel Pool Storage Racks ALL 4.
: 7. GL 93-04   Rod Control System Failure And Withdrawal Of Rod Control Cluster     WESTINGHOUSE
GL 95-07 Pressure Locking And Thermal Binding Of Safety. Related Power.
,                              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)
ALL Operated Gate Valves 5.
: 9. GL 92 01   Reactor Vessel Structural Integrity                                   ALL
GL 95-03 Circumferential Cracking Of Steam Generator Tubes PWR 6.
: 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
GL 94-03 Intergranular Stress Corrosion Cracking Of Core Shrouds In Boiling BWR l
: 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
Water Reactors 7.
: 14. GL 8814     Instrument Air Supply System Problems Affecting Safety Related       ALL Equipment
GL 93-04 Rod Control System Failure And Withdrawal Of Rod Control Cluster WESTINGHOUSE Assemblies,10 CFR 50.54(F) l 8.
: 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
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.
: 17. GL 8120     Safety Concerns Associated With Pipe Breaks In The BWR Scram         BWR System
GL 92 01 Reactor Vessel Structural Integrity ALL 10.
: 18. GL 81 14   Seismic Qualification Of Auxiliary Feedwater Systems                 PWR
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.
: 19. GL 81 12   Fire Protection Rule (45 F/R 76602, November 19,1980)                 ALL (Licensed prior to 1/1/79)
GL 89 21 Request For Information Concerning Status OfImplementation Of ALL Unresolved Safety Issue (USI) Requirements 12.
: 20. GL 8107     Control Of Heavy leads                                               ALL
GL 89-10 Safety Related (1) Motor-Operated Valve Testing And Surveillance ALL l
: 21. GL 78 09   Multiple Subsequent Actuations Of Safety / Relief Valves Following An BWR Isolation Event 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 17.
GL 8120 Safety Concerns Associated With Pipe Breaks In The BWR Scram BWR System 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
l l
l l


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        . 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 -
. Project No. 689
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Mr. Ralph Beedle Ms. Lynnette Hendricks, Director Senior Vice President Plant Support and Chief Nuclear Omcer -
Nuclear Energy Institute Nuclear Energy Institute _
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Washington, DC 20006-3708 j
Nuclear Energy Institute Suite 400 1776 i Street, NW .
Mr. Alex Marion, Director l
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
Programs' Nuclear Energy Institute Suite 400 1776 i Street, NW.
                . Nuclear Energy institute L
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 -
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
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 -
' Mr. Jim Davis, Director Operations Nuclear Energy Institute Suite 400 -
                - 1776 l Street, NW                                 .
- 1776 l Street, NW Washington, DC 20006-3708:
Washington, DC 20006-3708:
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Latest revision as of 07:35, 3 December 2024

Summary of 980423 Meeting W/Nuclear Energy Inst in Rockville,Md Re 10CFR50.59 Scope Issues.List of Attendees, Agenda Used for Meeting & Presentation Matl Encl
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)
v • d • e


Text

_

<> M ou g

UNITED STATES g

j 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 Regulatory Co (NRC's) offices in Rockville, Maryland, between representatives of the NRC N on's interested parties. Attachment 1 provides a list of attendees at the meeting A er includes the agenda that was used for the meeting and the presentation m n

NEl for the meeting. Attachments 3 and 4 are supplementalinformation pro e

y discussion of " safety analysis."

orthe three other topics related to the Commission SRM of March 2 sed the staff to reconcile their draft guidance document (NEl Conceming guidance for letter guidance as soon as possible. The NRR staff members stated that such d generic draft GL (sent to the Commission on April 20,1998).would occur o ons increases in radiological consequences. NEl continues an unreviewed safety question (USQ) if the acceptance limit (such as the P not be 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 that the li the facility design and operation. However, the degree of margin remaining to the ove be less as viewed by the staff than by the licensee. Therefore, if a license might changes that would have the effect ofincreasing calculated doses up to the j

e that the staff conclusion would be that the limits were actually exceeded NEl s

/

values could be allowed without always requiring NRC g^,~dg a

s 7,

9805070232 980504

/

PDR REVOP ERGNUMRC O Y /' "

Mf1 PDR

,d4 j j'>yQ g g

ya

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 ), 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 ). 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 ' s part of upcoming rulemaking on 10 CFR 50.59 criteria, the NRC a

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 -

SC:PGEBM7 OFFICE:

PM:PGEB PGEB a PWen:swh EMcKeb$N '

FAkstuhE NAME 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

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* '

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

= Enforcement policy change should be instituted before July 10 i

Y' 3

Purpose of 50.59

= Require licensee review of proposed changes a Determine if change exceeds previously approved design or operational limits

= Require prior NRC approval if any i

authorized limit is exceeded

'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 or Operating License?

Stop Yes P

Seek Amendment to No Order per 10 CFR 2.202 or Orders?

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 Apply NEl y,,

Change to Commitment Commitments?

Management Guideline Proceed With Change Update hSAR No No Regulatory per 10 CFR 50.71(e)

Interaction Required

i 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 How would it work?

= Identify safety analyses

. 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:

analyses included in the FSAR and approved by the Commission as part ofinitial licensing analyses performed pursuant to new or amended Commission regulations 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 l

DRAFT Identification of Safety Analyses i

Safety Analysis Basis for NRC Safety Analysis SER or other NRC Approval Requirement '

Reference 1.

General GDC 2.

Decrease in FW Temperature GDC 10,15,26 3.

Increase in FW Flow GDC 10,15,26 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 10.

Loss of Non-emergency AC Power to GDC 10,15,26 the Station Auxiliaries i

11.

Loss of Normal FW Flow GDC 10,15,26 12.

FW System Pipe Breaks inside and GDC 27,28,31 Outside Containment (PWR) 35,10 CFR 100 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, 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 23.

Inadvertent Operation of ECCS GDC 10,15,26 24.

CVCS Malfunction that increases GDC 10,15,26 Reactor Coolant inventory (PWR) 25.

Inadvertent Opening of a FWR Pr.

GDC 10,15,26 Relief Valve or a BWR Relief Valve 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)

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.

Radioactive Gas Waste System Leak or Failure 31.

Postulated Radioactive Release due to GDC 60, Liquid-Containing Tank Failures 10 CFR 20 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 36.

Temperature Effects on PWR Level IEB 79-21 Measurements 37.

Analysis of a PWR MSL Break with IEB 80-04 j

Continued Feedwater Addition i

38.

MOV CMFs during Transients due to IEB 85-03 Improper Switch Settings 39.

Pressurizer Surge Line Thermal IEB 88-11 Stratification in PWRs 40.

Seismic Qualification Of Auxiliary GL 81-14 Feedwater Systems l

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 44.

Equipment Operability / Containment GL 96-06 Integrity under DBA Conditions 45.

Assurance of Sufficient NPSH for ECC GL 97-04 and Containment Heat Removal Pumps 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 1.

IEB 96-03

, Potential Plugging of Emergency Core Cooling Suction Strainers by BWR Debris in BWRs 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 4.

IEB 93-02 Debris Plurring of ECCS Suction Strainers ALL 5.

IEB 90-02 Lose Of Thermal Marrin Caused By Channel Box Bow BWR 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 9.

IEB 88-07 Power Oscillations In Boiling Water Reactors (BWR)

BWR 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 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 21.

IEB 80-07 BWR Jet Pump Assembly Failure BWR 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 24.

IEB 79-21

, Temperature Effects On Level Measurements PWR 25.

IEB 7914 Seismic Analysis For As-Built Safety Related Piping Systems PWR 26.

IEB 79-12 Short Period Scrams At BWR Facilities BWR 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 17.

GL 8120 Safety Concerns Associated With Pipe Breaks In The BWR Scram BWR System 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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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

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