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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2066H001
        \. o./                                                January 28, 1998 MEETING SPONSORS: NUCLEAR REGULATORY COMMISSION (NRC) AND NUCLEAR ENERGY INSTITUTE (NEI)
 
==SUBJECT:==
INDUSTRY WORKSHOP ON GENERIC LETTER (GL) 96-06,
                                          " ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN BASIS ACCIDENT CONDITIONS" On Decert ber 4,1997, the NRC and NEl jointly sponsored an industry workshop on GL 96-06 in order to facilitate industry resolution of the GL 96-0G issues. The workshop was attended by approximately 90 people and was a good exchange of information between NRC and industry.
Attachment 1 contains a list of workshop attendeee. Presentations were made by NRC staff, NEl, licensee regesentatives, the Electric Power Research Institute (EPRI), and industry consultants. The moming session of the workshop focused on the waterhammer and two-phase flow issues, while the aftemoon session focused on thermal overpressurization of piping systems that penetrate containment. Attachment 2 contains the handouts and overhead slides that were used by both the NRC and industry during the workshop, including a copy of the agenda for the day.
Mr. Ledyard B. Marsh, Chief of the Plant Systems Branch, made opening comments for the NRC. He gave a brief history of GL 96-06, which included an explanation of why it was issued on an urgent basis. Several events occurred in 1996, including two plant shutdowns, which prompted the NRC to issue GL 96-06 quickly to ensure that licensees took the appropriate actions regarding operability of safety-related systems. As a result of work that has been done in response to the generic letter, licensees have completed operability determinations, and the staff now has some level of confidence that .*vstems are operable. Supplement 1 to GL 96-06 was issued on November 13,1997, to clarify the staff's expectations with respect to corrective actions for ti;e GL 96-06 issues and allow licensees time to reflect and ensure that the appropriate actions are being taken to address these issues. However, during his presentation, Mr. Marsh stressed that licensees should not infer from GL 96-06 Supplement 1 that they should not be concemed about operability and compliance aspects of affected systems and components. The safety concems that prompted GL 96-06 are still valid, represent challenges to safety systems, and should be appropriately dealt with. If a licensee has determined through        I evaluations that a particular corrective action is appropriate, these plans should not be delayed or modified.                                                                                              /
The staff discussed the status, findings, and lessons leamed from ongoing review efforts and answered questions from the industry. A summary of questions asked by the industry at the workshop and the staff's answers to those questions is contained in Attachment 3. The staff also discussed Supplement 1 to GL 96-06 and Revision 1 tn GL 91-18 ("Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability") regarding schedular considerations, corrective actions, and USQ (unreviewed safety question) considerations. The staff indicated that a Supplement 2 to GL 96-06 was being considered to address waterhammer concems gb1 K i
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.-    o-2                    January 28, 1998 associated with the scoping analysis for station blackout and to address concerns associated with waterhammer in other systems (e.g., RHR (residual heat removal) and spent fuel pool
  - cooling).
Presentations were made by industry representatives conceming their experiences, difficulties, and lessons learned in addressing the GL 96-06 !ssues. The status of EPRI testing and American Society of Mechanical Enginears (ASME) Code activities associated with the thermal overpressurization issue was also discussed during the workshop. EPRI provided the results of its pipe tests and presented a proposed analytical model to evaluate the maximum hoop strain in the piping resulting from thermally induced overpressurization.
Some of the major po:nts discussed during the workshop included GL 91-18, Revision 1, and USQ considerations, reasons that are (and are not) acceptable for delaying final resolution of the GL 96-06 issues (as allowed by Supplement 1 of GL 96-06), use of the ASME Code Appendix F criteria for permanent resolution of t.1e thermal overpressurization issue, content and schedule information about requests for additional information that will be issued by the NRC, and the overall schedule for completing review of the GL 96-06 issues. The staff indicated that requests for additionalinformation should be issued by the end of spring 1998, and the GL 96-06 reviews are currently scheduled to be completed by December 31,1999.
ls h, Beth A. Wetzel, Senio Project Manager Project Directorate lil-1 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation Attachments: 1: List of attendees 2: Presentation materials 3: Questions and answers
 
                                                                          +e    .
2                                    January 28, 1998 associated with the scoping analysis for station blackout and to address concerns anociated
'          with waterhammer in other systems (e.g., RHR (residual heat removal) and spent fuel pool cooling).
Presentations were made by industry representatives conceming their experiences, difficulties, and lessons learned in addressing the GL 96-06 issues. The status of EPRI testing and American Society of Mechanical Engineers (ASME) Code activities associated with the thermal overpressurization issue was also discussed during the workshop, EPRI provided the results of its pipe tests and presented a proposed analytical model to evaluate the maximum hoop strain in the piping resulting from thermally induced overpressurization.
Some of the major points discussed during the workshop included GL 91-18, Revision 1, and USQ considerations, reasons that are (and are not) acceptable for delaying final resolution of the GL 96-06 issues (as allowed by Supplement 1 of GL 96-06), use of the ASME Code Appendix F criteria for permanent resolution of the thermal overpressurization issue, content and schedule information about requests for additional information that will be issued by the i          NRC, and the overall schedule for completing review of the GL 96-06 issues. The staff .
Indicated that requests for additionalinformation should be issued by the end of spring 1998, and the GL 96-06 reviews are current!p scheduled to be completed by December 31,1999.
1 Original signed by Beth A. Wetzel, Senior Project Manager Project Directorate i .-1 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation Attachments:                  1: List of attendees 2: Presentation materials 3: Quer,tions and answers DISTRIBUTION:
See attached list 4
i        :3 DOCUMENT NAME: G:\WPDOCS\GL9606\DECMTG. SUM -
To receive a copy of this document, indicate in the box C= Copy w/o attachment / enclosure E= Copy with attachme7t/ enclosure N = No copy
(  OFFICE      LPM:PD31        E  t LA:PD31            E          BC:SPLEl      BC:EMEB e          D:PD31        C NAME        BWetzel:dtk              CJamerson C                  LMarsh        RWesNY            CCarpente DATE        01/2(,/98                01/ O /98 b 01/ N /98                        01/ y /98          01/ 1.8 /98 OFFICIAL RECOR COPY T$ g
 
b ,                                    . ...
DISTRIBUTION FOR NRC/NEl DECEMBER 4.1997. MEETING ON GL 96-06 DATE: January 28, 1998 Hard Coov Central File PUBLIC PD# 3-1 Reading B. Wetzel OGC ACRS NEl Operating reactor licensees Todd Baccn Duke Engineering 215 Shuman Naperville,IL 60563 Timothy S. Andreychek Westinghouse Electric Corp.
P.O. Box 355 Pittsburgh, PA 15230 E-Mail S. Collins /F. Miraglia (SJC1/FJM)
R. Zimmerman (RPZ)
E. Adensam (EGA1)
C. Carpenter (CAC)
C. Jamerson (CAJ1)
B. McCabe (BCM)
T. Martin (SLM3)
OPA (OPA)
G. Tracy (GMT)
D. Screnci (DPS)
M. Tschiltz (MDT)
T. Hiltz (TGH)
K. Clark (KMC2)
J. Fair (JRF)
G. Hammer (CGH)
J. Strasma (RJS2)
J. Tatum (JET 1)
R. Wessman (RHW)
B. Henderson (BWH)
L. Marsh (LBM)
G. Hubbard (GTH)
C. Saadu (CYY)
J. Guzman (JGG1)
K. Manoly (KAM)
: 1. Jung (IXJ) m
 
_. _..      .  ~. ._                                          -
  . a JOINT NRC/NEl INDUSTRY WORKSHOP GL 96-06 December 4.1997 Name                                  Affiliation
: 1. Kevin Browning                        Penna. Power & Light
: 2. E. Forest                              NVS LIS
: 3. Vaughn Wajoner                        CP&L 4  Kenneth Canavan                        GPU Nuclear
: 5. Paul Hirschberg                        Pacific Gas & Electric Co.
;  6. Edward A. Wais                        Wais & Associates
: 7. Steve Gosselin                        EPRI
: 8. Steve Thomas                          NSP
: 9. Quoc Huynh                            South Texas Project (STP)
;  10. Lewis Allen                            STP
,  11. Tom Esselman                          Altran Corp.
: 12. Greg Zysk                              Altran Corp.
: 13. Sontra Yim                            Altran Corp.
: 14. Jack Hamm                              FPL 15  Subhash Klusana                        FPL
: 16. Ahmad Shahrpass                        NMPC
: 17. Robert Campbell                        TVA
: 18. Bruce Heida                            WPSC/KNPP
: 19. Todd Bacon                            Duke Engineering
: 20. Greg Krueger                          PECO Energy
: 21. Mark Msaatvedt                        PP&L
: 22. David Nix                              Duke Energy /ONS
: 23. Kalyan K. Niyogi                      Holtec International
: 24. Kevin Borton                          PECO Energy
,  25. Jorge E. del Mazo                      Pacific Gas & Electric Co.
: 26. Bob Coward                            MPR Associates
: 27. Chris Ludlow                          BGE
: 28. Eric May                              Virginia Power
: 29. Q. Shane Lies                          American Electric Power
: 30. Tom Ryan                              Sargent & Lundy
: 31. Bob Jones                              NSP
: 32. Skip Denny                            PECO
: 33. Tim McClure                            Cooper Nuclear Station i
: 34. Beth A. Wetzel                        USNRC, Lead PM, GL 96-06
: 35. John Fair                              USNRC, Mechanical Engineering Br.
: 36. Gary Hammer                            USNRC, Mechanical Engineering Br.
: 37. Kurt Cozens                            NEl
: 38. Jim Tatum                              NRC/SPLB
,  39. Dick'Wessman                          NRC/EMEB                                                      <
: 40. L.B. (Tad) Marsh                      NRC/SPLB
: 41. George Hubbard                        NRC/SPLB ATTACHMENT 1 d
 
        . o JOINT NRC/NEl INDUSTRY WORESHOP GL 9646 (Continued)
Name                                    Affiliation
: 42. Coretta Saadu                            NRC/SPLB
: 43. Dave Modeen                              NEl
: 44. Kim Hull                                WPSC/KNPP
: 45. Joel Guzman                              USNRC, Rlli
: 46. Steve Reckford                          BGE Calvert Cliffs
          '<  Phil Fienner                            Palisades
:t  Bill Peebles                            Sargent & Lundy
: 49. Bud Gerling                              Coasumers Energy - Palisades                '
: 50. Cory Flensburg                          First Energy - Perry
: 51. Todd Conner                              Ealtimore Gas & Electric
: 52. Bob Henry                                Fauske & Assoc , Inc,
: 53. Keshab Dwivedy                          Virginia Power
: 54. David Murphy                            Bechtel
: 55. R.T. Thw6a"                              Entergy
: 56. Allen Vieira                              Bechtel '
: 57. Kamal Manoly                            NRC
: 58. Daniel A. VanDuyne                        Stone & Webster (Northeast)
: 59. Robert G. Vasey                          AEP
: 60. Hossein P. Nourbakhsh                    BNL
: 61. Paul G. Schoepf                          AEP
: 62. Jeff D. Heminger                          Duke Energy /Oconee
: 63. Jason A.- Patterson                        Duke Energy /Oconee
: 64. Dave Shafer                              Union Electric /Callaway
: 65. Bruce Deombls                              Virginia Power
: 66. Mike Henig                                Virginia Power
: 67. Rick Ingram                              Entergy
: 68. Bob Hammersely                            FAJ
: 69. Joe Paljug                                FTl
: 70. Avtar Singh                                EPRI                                      -
: 71. Timothy S. Andreychek                      Westinghouse
: 72. Kenneth A. Hudson                          F2CO Nuclear
: 73. Raubin Randels                            Comed
: 74. Steven Unikewicz                          Millstone Nuclear Station
: 75. Paul Rainey                                Vermont Yankee
: 76. Khalil Shaupbi                            Con Ed 77  Altheia Wyche                              SERCH Licensing /Bechtel
: 78. L.T. Tang                                  EPRI
: 79. G.S. Shukla                                Detroit Edison
: 80. Philip L. Walker                          South Texas Project Nuclear Operating Co.
: 81. Gil Williams                              South Carolina Electric & Gas Co.
: 82. Ian L. Jung                                NRC/DSSA/SPSB
: 83. Bill Selbe                                Wolf Creek I  .-.
 
l a                                                                                                                                              1 1
4 NRC AND NEl JOINT WORKSHOP
,                                                                                                          ON GENERIC LETTER 96-06 4                                                                                                    DECEMBER 4,1997 J
T i
f 1.
i i
s ATTACHMENT 2
 
NRC AND NEI JOINT WORESHOP ON GENERIC LETTER 96-06 TAE1  LOGISTICS
                . Meeting Agenda TAB 2  OVERVIEW AND GL 96-06 STATUS                  -
                . NRC Presentations on Background, Characterization and Status of Reviews                                                      .
TAB 3  SUPPLEMENTS TO GL 96-06
* NRC Presentations on GL 6-06, Supplement 1 and GL 91 18, Revision 1
                . NRC Presentations on Waterhrumer and Two-Phase Flow Findings / Lessons Learned (Considerations for analytical solutions, Potential Supplement 2) h3 4  INDUSTRY EXPERIENCE WITH WATERHAMus'R ISSUES
                . Consumers Power Presentation on Palisades
                . Northeast Utilities Presentation on Millstone Unit 2
                . Baltimore Gas and Electric Presentation on Calvert Cliffs TAB 5  THERMAL OVEEPRESSURIZATION ISSUES
                  . NRC Presentation on GSI-150, Use cf Appendix F, Code Inquires, and I
the Proposed ASME Code Case TAB 6  OVERPRESSURIZATION TEST PROGRAM t
                  . EPRI Prevention on Overpressurization Test Program and Results TAB 7  INDUSTRY EXPERIENCE WITH THERMAL OVERPRESSURIZATION ISSUES
                  . Wolf Creek Presentation
                  . Northeast Utilities Presentation on Millstone Unit 2
                  . Florida Power and Light Presentation on Turkey Point TAB 8  RISE / RELIT 2ILITYINSIGHTS
                  . NRC Presentation
                  . PECO Presentation
                  . GPU Presentation TAB 9  REFERENCE INFORMATION l
 
AGENDA JOINT NRC/ INDUSTRY WORKS!iOP ON GL 96-06 December 4,1997 Time            Topic                                                        Presenter Mornine Session 8:00      Opening Remarks                                                  NRC and NEl 8:15      Background, Characterization of issues and Status of Rev!sws      NRC 8:40      GL 96-06 Supplement 1 and GL 91-18 Rev.1                          NRC 9:00      Waterhammer and Two-Phase Flow Findings / Lessons Learned - NRC (Considerations for analytical solutions, Potential Supplement 2) 9:45      Break 10:00      Industry Experience with Waterhammer issue            .        Consumers Power NU BG&E 12:00      Lunch Afternoon Session 1:00      Opening Remarks                                                  NRC and NEl 1:15      Thermal Overpressurization issues                                NRL
* GSI-150
                  . Use of Appendix F
                  . CcJe Inquiries
                  . Proposed Code Case 2:00      Piping Overpressurization Test Results                          EPRI 3:00      Break 3:15      Industry Experience with Thermal Overpressurization issue        Wolf Creek NU FPL 3:45      Risk / Reliability Insights                                      NRC PECO GPLI 4:30        Summary Remarks and Future Activities                          NRC and'IEl 5:00      Adjoum
 
O 2
                                                          ~
NRC INTRODUCTORY REMARKS r
                                          $gc.,kBREGQ*o g,
4 ,,
4....
                                                    /
Presented By: L. B. (Tad) Marsh, Branch Chief Plant Systems Branch Office of Nuclear Reactor Regulation
 
                                                                            -11
              + HISTORY OF GL 96-06 DEVELOPMENT                              -
Compliance emphasis based on Millstone, Maine Yankee and Haddam Neck Events / Information -
Diablo Canyon SW modification Westinghouse Nuclear Safety Advisory Letter Haddam Neck shutdown due to inoperable CFU's i
Maine Yankee shutdown due to thermal                    i overpressure concerns GL 96-06 developed - fast track without public comment Industry concerns communicated informally to NRC            ,
1 t
 
!        +  SAFETY. AND COMPLIANCE l            -                  COMSAJ 97-008 articulates Commission position on
:                              safety and compliance, and on using risk insights in l                              regulatorf decisions
        +  GENERIC LETTER 96-06 SUPPLEMENT 1
* Clarifies staff expectations on schedules l
* Allows modification of initial 96-06 corrective actions schedule I
            -                  Use risk insights i
i l
2 l                                                                                                        ,              ~
i l
 
O
          +      NRC REMAINS CONCERNED ABOUT GL 96-06 ISSUES                                      -l No reduction in agency concern about basic issues Challenges to safety systerns i
                -                Licensing basis concerns i
3
 
BACKGROUND,
                                                                                ~
CHARACTERIZATION OF ISSUES, AND REVIFW STATUS Presented By: Coretta Saadu Plant Systems Branch Office of Nuclear Reactor Regulation
 
REQUESTED INFORMATION
                                                  ~
+ GL ISSUED SEPTEMBER 30,1996
+ 120 DAY RESPONSES DUE JANUARY 28,1997            l
  =    If Susceptible
* Actions Taken
  . Basis for Operability
  -  Corrective Actions 1
 
GL RESPONSE.
 
==SUMMARY==
 
    + 7 PLANTS NOT SUSCEPTIBLE TO ANY OF THE ISSUES
    + 61 PLANTS SUSCEPTIBLE TO WATERHAMMER AND/OR TWO PHASE FLOW
    + 97 PLANTS SUSCEPTIBLE TO THERMAL.
OVERPRESSURIZATION
    + 28 PLANTS SUSCEPTIBLE TO ALL THREE ISSUES l
 
CORRECTIVE ACTIONS                                      -
i
    + WATERHAMMER / TWO PHASE FLOW
      + Performed Calculations
      - Revised Emergency Operating Procedures
      - Installation of Surge Volume
      + Increase Nitrogen Pressure in the Component Cooling Water System Surge Tank 3
 
GL 96-06, SUPPLEMFNT 1
                " ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS" GL 91-18, REVISION 1                  .
1 "INFORMATION TO LICENSEES REGARDING NRC INSPECTION MANUAL ON RESOLUTION OF DEGRADED AND NONCONFORMING CONDIT!ONS" Presented By:  George Hubbard, Section Chief Plant Systems Branch Office of Nuclear Reactor Regulation
 
b GENERIC LETTER 96-06 SUPPLEMENT 1
  +  PURPOSE
* Ongoing Efforts
* New Developments
* Additional Guidance for Completing Corrective Actions 1
 
4 CORRECTIVE ACTION GUIDANCE                          j t
                                  + LICENSEES RESPONSIBLE FOR:
Assessing Operability
* Determining Actions                                      :
* Establishing Schedules                                  j
                                  + CONSIDERATIONS
;                                  -  Validity of Operability Determinations
                                    -  Compensatory Actions for Operability
                                    +  Safety Significance
* Risk insights                                        '
                                    -  Time for Completing Generic initiatives and/or Plant    j Specific Actions
 
CORRECTIVE ACTION GUIDANCE                    '
(cont.)
                + ANALYTICAL SOLUTIONS FOR PERMANENT RESOLUTION
* ASME Code, Section Ill, Appendix F
* Other Acceptance Criteria
                  -  Viable if Appropriate, Justified, and Evaluated in Accordance with NRC Requirements (e.g.,10 CFR 50.59) s
. _ _ _ _ _ _          -                        .        ~
 
CORRECTIVE ACTION GUIDANCE (Cont.)
:                                                                                                                                                      1 I
;    +      GENERIC LETTER 9148, REVISION 1, DATED OCTOBER 8,1997                                                                                                                                ;
    +    SCHEDULER ADJUSTMENT MUST BE JIJSTIFIED AND SHOULD BE COMMUNICATED
    +      ASME CODE,.SECTION lli, APPENDIX F, FOR INTERIM OPERABILITY Until Permanent Action is identified and Approved by NRC das applicable)
          -                                  Supplements GL 91-18 for GL 96-06 issues 4
 
                                                                        ~'
,                      GENERIC LETTER 91-18 REV.1
      +    PURPOSE i
* Inform Addressees of Revised Section of Part 9900,
                  " Technical Guidance (' Resolution of Degraded and Nonconforming Conditions')"
Does Not Revise Part 9900.Section " Operable /
.                Operability: Ensuring the Functional Capability of a System or Component" 1                Discusses Role of 10 CFR 50.59 in the Resolution of Degraded and Nonconforming Conditions Guidance on Effect of Unreviewed Safety Questions
;                (USQs) on Plant Operations                                ,
5
 
PART 9900: TECHNICAL GUIDANCE
                        " RESOLUTION OF DEGRADED AND NONCONFORMING CONDITIONS"
  + SECTION 4.3. THE CURRENT LICENSING BASIS AND 10 CFR 50, APPENDIX B When Corrective Action Starts -- Discovery Considerations for Prompt Corrective Action
;                        Safety Significance
,                        Actione Necessary to implement Corrective Actions i
e n,
 
        " RESOLUTION OF DEGRADED AND NONCONFORMING CONDITIONS" (cont.)
  + Section 4.3 (cont.)
* Factors influencing Time for Corrective Action Design, Review, and Approval Procurement of Repair / Modification Availability of Specialized Equipment Need for Hot or Cold Shutdown
    - Time Frames Longer Than the Next Outage Should Be Explicitly Justified By the Licensee l
i c i-e m---
 
l        " RESOLUTION OF DEGRADED AND NONCONFORMING CONDITIONS" (Cont.)
i
  + SECTION 4.7: EVALUATION OF COMPENSATORY MEASURES l
* Operability Determinations
!        Interim Compensatory Measures                          .
Operator Action
* Resolution Should Be Prompt
                                                              ~
8
 
      " RESOLUTION 0F DEGRADED AND NONCONFORMING CONDITIONS" (cont.)                          '
+ SECTION 4.8: FINAL CORRECTIVE ACTION
* Need for NRC Approval (e.g., USQ) Does Not Affect Licensee's Authority to Operate the Plant Provided:
Necessary Equipment is Operable Degraded Condition is Not in Conflict with T.S.
D
 
==SUMMARY==
 
      + NRC IS RECEPTIVE TO CHANGES IN SCHEDULE FOR GL 96-06 CORRECTIVE ACTI.ON AS LONG AS JUSTIFIED
      + ASME CODE, SECTION lli, APPENDIX F, MAY BE A VIABLE PERMANENT SOLUTION
      +  ASME CODE, SECTION lil, APPENDIX F, MAY BE USED FOR INTERIM OPERABILITY UNTIL PERMANENT ACTIONS HAVE BEEN IDENTIFIED AND APPROVED BY l
NRC (AS APPLICABLE) l i
10
 
1
!                                                            I l                             
 
==SUMMARY==
                  ~!
(cont.)                    !
i i
j  + IF SCHEDULER CHANGES ARE MADE A REVISED l    RESPONSE TO GL 96-06 SHOULD BE SUBMITTED
  + GL 91-18 REVISION 1 RESOLVES USQ ISSUE i
f I
i
!                                    11 i
 
                                                                                      'l
                                                                                      .i WAT_ERHAMMER AND TWO-PHASE FLOW l
    +      NRC REVIEW CONSIDERATIONS
    +      FINDINGS / LESSONS LEARNED
    +      OTHER SYSTEMS AFFECTED
)
i I
k Presented By: James Tatum Plant Systems Branch Office of Nuclear Reactor Regulation
 
i                                                        .
;            NRC REVIEW CONSIDERATIONS                    ~!
i
!  + RAls WILL BE ISSUED (focus will be on methodology, j    assumptions, justification, and corrective actions) i i  + SEs WILL BE WRITTEN 4
i
    + CONTRACT SUPPORT WILL BE OBTAINED AS NEEDED i
i i                                                          !
!                                                          i t
 
NRC REVIEW CONSIDERATIONS (cont.)
  +                        ANALYTICAL METHODOLOGY i                          -
CLASSICAL APPROACH
;                              (NUREGICR-5220)
!                            ALTERNATE          APPROACH (BASIS / JUSTIFICATION) l                                  Applicability j                                  Test Data (Good and Bad)
Uncertainty Analysis i
I                                                                                          ~
l                                                          2 i
 
                                                              ~
NRC REVIEW CONSIDERATIONS (cont.)
  +  ANALYTICAL METHODOLOGY (cont.)
Assumptions and input Parameters Including Justification for Plant-Specific Application (e.g., the use of engineering judgemant must be supported by research and test data applicable to the specific system design and configuration)
Cushioning Speed of Sound Force Reductions ALL SCENARIOS & WATERHAMMER TYPES (identify worst case) 3
 
mes, NRC REVIEW CONSID_ERATIONS l                          (cont.)                            ;
1 l
l
  + ANALYTICAL METHODOLOGY (cont.)
    - Amplification Due to Fluid-Structure Interaction Licensing Basis Envelope (e.g., worst-case temperatures, pressures, flow rates, load combinations, and single failure) 4
: u. - -. -...m.._ . _ . _ . ,  ..u-    4---
                                                      *a-i----..&+ - - . * = -    '-a = ~ sma--=~4--+---+m'~-'-"'--w-=-----=  4=u=-a--== s*-  ==-= -++ - -~*~ =-- - - - -~= = - --- ------ + = - --=---au-m.~        ----*
4 4
NRC REVIEW CONSIDERATIONS
                                                                                                                                                                                                                          ~
;                                                                                                            (cont)
I i
            +                  ANALYTICAL METHODOLOGY (cont.)
l l
Two-Phase Flow Effects of Void Fraction on Flow Balance and j                                                      Heat Transfer Consequences of Steam Formation, Transport, and Accumulation I
Cavitation, Resonance, and Fatigue Effects Erosion Considerations NUREGICR-6031, " Cavitation Guide for Control I                                                    Valves" 5
 
i 1
l NRC REVIEW CONSIDERATIONS          !
l l                          (cont.)            t I
  +  MODIFICATIONS 1
i USQ Determination j  -
Technical Specification Considerations Adequacy of Modification t
l
  + ACCEPT WATERHAMMER AND TWO-PHASE FLOW      '
CONDITIONS i
Impact on Other Evaluations l
I i
6              'I
 
l FINDINGS / LESSONS LEARNED                                      -l
!                                                                                        i
  +    BASIS FOR ANALYTICAL ASSUMPTIONS NOT PROVIDED 3              Cushioning                                                              i Use of Force Reduction Factor Reduction in Acoustic Velocity i
:  +  INCOMPLETE ANALYSES                                                              :
;                                                                                        I 4
i l
i      -
Inadequate Single Failure Analysis l
Systems / Structure Amplification Not Addressed Effects of Two-Phase Flow Not Addressed                                  ,
I 7
 
I FINDINGS / LESSONS LEARNED l                                                                                              (cont.)
i 4
l  + OTHER ANALYSES MAY BE AFFECTED i
EQ
    -                                              Station Blackout
!  + SOME MODIFICATIONS MAY REQUIRE NRC REVIEW AND
:    APPROVAL
  + RELATED AMENDMENT APPLICATIONS l                                                                                                  8 I
 
OTHER SYSTEMS AFFECTED                          '
+  OTHER SYSTEMS MAY BE AFFECTED BY WATERHAMMER (primarily column separation and rejoining)
IN 87-10, Supplement 1, " Potential for Water Hammer During Restart of Residual Heat Removal Pumps,"
dated May 15,1997
  . IN 91-50, Supplement 1 " Water Hammer Events Since 1991," dated July 17,1997
  -  Examples:        RHR; Spent Fuel Pool Cooling; Recirculation Spray System; and Low Pressure Safety injection 9
 
OTHER SYSTEMS AFFECTED (cont.)
+    GL 96-06, SUPPLEMENT 2, UNDER CONSIDERATION 10
                                                                                                        . . . . . . _  i
                                                                      ._ =
 
==SUMMARY==
!        +  NRC REVIEW IS ONGOING AND RAls WILL BE l            FORTHCOMING l        +  NRC REVIEW WILL FOCUS PRIMARILY ON:
Analytical Assumptions and Inputs                                              i I
Adequacy of Modifications i
3 Ultimate Resolution of the issues
          +  SUPPLEMENT 2 TO GL 96-06 IS UNDER CONSIDERATION l
i 1                                                                          11 i
 
i l                                                                                          1 l
Potential Waterhammer in Service Water System Palisades Plant Consumers Energy Sargent & Lundy Fauske & Associates
                                          - - - - - ~ . - - - -
 
  , hat . .r .J.4__4      A >      , 244 y- e-+a --6h2-,.+,p4- 4E--h._ . - -A E4----J -
m-5-r-d=.A      4 MSm-.w sh -h..,MM4 -m e                        Cae.is46 M.4 ACAMJ 45 A AJ.XhM--bM*44&-f"M-4                    M-44 i
                    -.e==                                                                                                                                                                                                                        -        i y
mamm i
i
              +
f i
i i
e i
i f
i I
1 4
2 I
i 4
O                                                                                                                                                              .
5 Du 3                                                                                                                                                              i o
k 4
U M
IWW
 
i i                                            i  -i Containment Air Cooler / Service Water Piping Configuration
      . Once-Through Open Service Water System
      . Lake Intake @ Elev. 579
      . Four Containment Air Coolers @ Elev. 600
      . Containment Service Water Penetrations @
Elev. 612 (Inlet and Outlet) .          .
      . Discharge @ Elev. 583
 
i I
i l
Operability Assessment
                            . Utilized Simple Thermal / Hydraulic Models
                            . Estimated Amount Of Steam Generation
                            . Tracked Location Of Steam Bubble
                            . Determined Boundary Conditions for Start of Refill Transient
                            . Performed Refill Transient Analysis (After Pump Restart)
                                + Estimate of Maximum Refill Velocity E
 
      .i                                                                        1 1
!            Operability Assessment (continued)
;          . Void Collapse Assumed In Piping Downstream Of Air Coolers
!          . Conservative Estimate Of Waterhammer                                      i
!            Pressure Determined Using NUREG/CR 5220
          . Piping Analysis Performed To Demonstrate
!            Operability                .
l i
4 i
                                                                              ,        i i  i                                                                              i j
j
 
4      ,t 4
w m
Ta
      %    ~
Ta Ta z
n D
k.
W h
I
 
    ,                                                    a Characterization ofRealistic Void Collapse Mechanism
      . Objective
        + Quantify a more mechanistic assessment of collapse of voided region.
        + Perform a scoping study to quantify margin in dynamic load magnitude during column rejoining.                  -              -
L                                                        E
 
I I
Characterization ofRealistic Void Collapse Mechanisin
                                . Approach
                                  + Assess dynamics of slug-bubble trapped between two liquid columns.
                                  + Initially, all are at rest.
                                  + Upon SW pump restart, one liquid column will begin to refill, push and compress the void.
e e
 
I                                                          l Characterization ofRealistic Void Collapse Mechanism
    . Approach (continued)
      + Void collapse rate is quantified by solving the momentum equations of the two surrounding water columns.
      + Effect of vapor condensation in void in the presence of inert gas is included. Inert gas comes out of solution when the water boiled.          .
O i
 
l i
Characterization ofRealistic Void Collapse Mechanism D                Slug
(
Bubble                    Receiver Supply                q Reservoir        .
Reservoir X%  3 column i          P  column 2 2
* pO L        PS                                                    y
(
X'    X2  =
q
                          )        ,=9 ME96D045.CDR I2-9-96 I
I
 
l
    ,                                                        ,  l Characterization ofRealistic Void Collapse Mechanisin
      . Assumptions
          + Non-condensable gas and vapor obey perfect gas law.
          + Temperature of liquid and pipe wall contacting bubble remain constant and equal initial liquid temperature.
          + Initialliquid temperature equals boiling point at internal pipe pressure.
b I
 
Characterization ofRealistic Void Collapse Mechanism
                                                    ~
* Assumptions (continued)
        + Rate of condensation is controlled by purely            :
turbulent mass transport within the bubble.
        + Effective friction loss coefficients account for obstacles to the flow of the two water columns.
        + Variations in the liquid column masses during the
;          void collapse transient are ignored.            -
!                                                                  I E
g e l
 
                ,                                                      i Characterization ofRealistic Void Collapse Mechanisin
                  . Model
                    + Three coupled differential conditions
                      - An equation of motion for each liquid column.
                      - An equation for mass of vapor (steam) within the bubble.
o
 
I                                                                                                                                      i Characterization ofRealistic Void Coi!i! apse Mechanisin                                                                                                  ;
MOdel (continued)                                                  -
                  + Initial conditions include:
                    - Location of interfaces between void and upstream and downstream water columns.
                    - Zero velocity for each column.
                    - Mechanical equilibrium such that the two reservoir pressures equal the bubble presstire.                                                                                  -
                  + At time = 0, supply pressure is increased to start void collapse.
R g
e
 
                                                                                    .I i                                                                  i Characterization ofRealistic Void Collapse Mechanism e  Sample Results
                            + Void collapse is gradual (not instantaneous) and dominated by vapor condensation for bubbles formed by boiling.
                            + Bubble pressurizes and decelerates upstream column.
                            + Downstream column is accelerated and moving when two columns rejoin.
 
l l                                                            l Characterization ofRealistic Void Collapse Mechanism l
      . Sample Results (continued)
        + A " soft" landing due to cushioning effect of bubble between two water columns and their limited relative velocities.
        + Overpressure due to column rejoining is generally small (10s of psi) for situations where the bubble is surrounded by high temperature water.
g I
a
 
I                                                                                                                                  I Characterization o Realistic Void Collapse Mechanism - Results 38    .........,.........g.........y.........g.........g.........g.........y.........y.........g.........
8      ~
13  -                tt * "8 T                                                                            -
            =
            .      -                Ea . see.e j      _                v, . so '
e      -
eJ O  33  .________________.A3__"'
e                                                                                                                                      .
          =
s e
          #                                                                                                                                      I
            .                                                                          *                                        ,              i 3    s  -
e n
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i
                    .  .....l.........l.........l.........l.e......l.........l.........l.........l.........l..........                          l 9            I      2          3          4        5            6        7        8          9      16 Ties, e                                              .                        l f                                                                                                                                        E    E I
 
1                                                                                                                                                                                                      I a
Characterization o Realistic Void Collapse Mechanism - Results                                                                                                                                        [
1.S      .
                                                                                        ........,.........g.........g..a......,.........g.........,.........g.........g.........g.........
I
_                                                                                                                                                              t k                                                                            -
2    -
a          -                                                                          Tf* ''' T                                                                          t
                                                                          .                                                                                    sg . em.e                                                                          .
I v, . so
* i  s.s    -                                                                                                                                                              .
2e M                                                                            #
a .. -                                                                                  i e                                                                      0 e
e    ,    .    .
                                                                                                                                              ,/                                                                                                  ,
e                                                                  l                                                                                                    i u                                                                  ,                                                                                                        ;
1                                                                l
: e.  .S        -                                                t                                                                                                        ,
m3                                                            0 8
9 9
                                                                              ,    .  ...un1            -        l------'f              l          *l      l        l      l          -
9              3        3      3        4            $            6      7      8        9          30 Time. o                                        .
I                                                                                                                                                                                                          I e
i.
 
l                                                                                                                                                          l Characterization o Realistic Void Collapse Mechanism - Results m    .s    -                                                                                                                          -
f          .
v, . ses c
        =
                    .                                                                                    s, - see.e s, - ee *
      ,#    .4 se M                                                                            ,
g .a.    =InsamaamaaIman_amamaInanaaaaaaInanaaaaamIa amamaamImmananaamInamaamaaaIsaaamamaalmamammaaa.
9 e          3        2                        3    4      5            6            7                  8          9      IS Time,    s 1
 
j I
i Integrated Thennal Hydraulic Analysis
                          . Transient Integral Assessment Of Fan Cooler And Attached. Piping Response To Pump Trip And Recovery 4
I e
 
Integrated Thermal Hydraulic Analysis
                          . Thermal-Hydraulic Phenomena Considered:
                                + Pump coast down,
                                + Two phase flow,
                                + Heat transfer (thermal response of piping and fan cooler),
                                + Void formation and steam condensation,
                                + Void collapse and water column rejoining, and
                                + Acoustic velocity including residual void.
I I
                                                                                    . , .. . . . -    =
4 .
 
I                                                      I Integrated Thermal Hydraulic Analysis
                .      Forcing Functions Based On Momentum Change In Pipe Segments l
l I
i
 
6 i
    .o 9
W rf3 4    .
o (ta
    +
e o
M tr>
4 umu
 
l Joint NRC/ Industry Workshop December 4,1997 Millstone Unit 2 Experience with GL 96-06 WaterhammerIssue by Kalyan K. Niyogi        .
H  O  LT  E  C INTER N ATION A L  -
l
 
Millstone Unit 2 Experience with GL 96-06 WaterhammerIssue
        ^
m introduction m System Description a Computer Modeling m Results              .
m Conclusions
 
===System Description===
m RBCCW System Characteristics a Design Conditions m System Alignments m Restart Sequences O
 
r            M3GE                              HEADER A SCHEMATIC                                                  EL53 TANK CEDM EL 80                                                                                              X-34B                                      l JL EL45            EL45        l CEDM            CEDM X-34C          X-34A      ;
EL43                                                                                    a              n      l CAR X-35A J L 10" EL15 RCP 10"                        EL4                  CLRS CAR                        a X-35C              3"            3"              3"              3" n
Ak 10" m
EL -4
                                                                          ' '    )k f                                                                                    16"                    6"    1'                      ,
j                                                              16"                                %      ;                4 20"                    EL -16                1 f                      y 8" EL-10                          6*                a SFP HX, SD HX ESF &Si PMPS h20"                                  k                                                "
3*
WASTE GAS COMP                                      16 RBCCW EL -16                16 j      f NC  8*                i EL-10 6"
                                                                                                                                                  ^
HX-18A        l    P SlP, CSP.ESF RM CLRS, SD HX EL-40                                                          4  .
EL -23        20            ---> SFP HX EL-40                                                                  EL-1g    QT RCP
                                              --> MSTE GAS COMP EL-20
 
Ii I      l    l1l u
6 k
4 "6      RD5                                  "
5  A3                                    0 L  C-X                                  1 E        '
* p                "0 5    CP 1    SC                  1 L    VR E    R& l 3-L G        E E      L DND 4
P. T                    "0 ME                      1 A
S L-4                        -
                                              .                          0 0                          1 0
1 4
L
                                                                                    "              E C
I                                                ,
6 1
0      X T                                                ,      8 2- H A                                                                                        L E S4 -
D    0 M                      *                                                                  . SL E                      0                              <                        0        NRE H                      1                                                      1        OL
                                                                <    - '              -      C C X C      3 4 R85                                              "
L E        TMH S      L A3                                                6 "0          NRH E C-X  k 1
1            E SN    FC B              4 k
4 g
j V
SEE R                                                            2 A P.U GSQ E                                                            1 2        EC X D P, H D                                                            L E
1 I
A                                                                              L E              SP F
E                                                                                                    S H                                                                              "
l i
                                                                      "            0
                                                                "                  2 0
2    f                          3 X                        1 H              C D            W8        L E
S      N. C1 -
S,      O    CX B-C    RHX RX N LH    T C  H E MC V              .
RN S FE A SU G EQ E P.  , D S X CH P, FP                3 I
2-SS
: e. 0 8
2 '                      L E
pK L RN E yA ST L
 
l Equipment Characteristics m Pumps a CAR Cooler Units a CEDM Cooler Units a Surge Tank WP O
 
CAR Cooler Units Physical Arrangement Performance Characteristics I
i t
I
 
a.a,A.-.- 4 4 4- ._ g.. g p_.4. m .h  b 4- e-.: --h-.'"''A-n.m*4--dE-EAF.a.*-E  .-.--aem wu,..w a --4  & E*'m h--  .m-    Ah.    *i+aw.k--ma    da a,ms4.hsa..g  mh-  ''m.
                                                                                                                                                                              --    A.-.*-.mhan.,da->      ea m h  Le** -  +A-=.%    ,a E-*-h-e  , -m- -'--a4ae+4- h2*.4& 5m u.w  kh--
d i
                                                                                                                                                                                                                                                                                          /
k OOLg4                                NkAy                              444****e,7 s
i o
                                                                                                                                                                                                                                                                                          /
                                                                                                                                                                                                                                                                                          /
i O
/
l 4
/
o                                                                                                                                                                              /
                                                                                  /                                                                                                                                                                                                        '
                                                                                                                                                                                                                                                                                          /
                                                                                                                                                                                                                                                                                          /
/
Q            0                        /
                                                                                                                                                                                                                                                                                          /
                                                                                                                                                                                                                                                                                          /
/                                                              d
/                                                                                                                                                                                                                                                                                          ,
i
                                                                                                                                                                                                                                      \
                                                                                                                                                                                                                            \
k                                                      !
                                                                                                                                                                                                                          \                      '
                                                                                                                                                                                                                        %                                                                /
s s
o'                                                                      '
ss oo s
j                                                                        o a
s s
0                                                                                o                                                                                                                  !
h
                                                                                        /                                                                    \
It I
sI
                                                                                                                                                    \
                                                                                                                                                \                                                                                .                                                        ,
                                                                                                                                          \                                                                                                                                        e g \                                                                                                                                                  /
g g
g                                                                                                                                                            i g
                                                                                                                                                                                                                                                                                    ~
 
t *
                                                                                                                                                              .+                                                                  Y
                                                                                                                                                            ,sf ''
l .            .:k ;
s              v
                                                                                                                                                                                                                      . l.
sW
                                                                                                                                                                                                                      ~
                                                                                                                                                                        , . <                                                                                                  s                            <
                                                                                                                                                                                                                                        "3 s                                                      /              c J.U:        . y              s                        ..                                                                                  .t              .
s:Qxsm9 "ys                                  ^
                                            .k..[              t        >
: fM[l j' "
                                                                                                                                                                        #                i                                        4  I :.R y                                  ^                                  I '
JM..
wg.. ;. ,, },
f ''(: . ,3 , ' , ,
                                                                                                                                                                                                                ;,                                    v                        ,b<          q..
                                          '['f ."<: 7                                                                                                    , }j ; .
s
                                            ,4,[$-                              s                ;m,                                                                                                        ,  ',,#                                                            s            s qq3
                                                                                                                                                                                                                                                                                                                                        ..f }'{%                  y n                                      <                                                                                s                                                                          e
                  .>.f. -.
                                                                                                                                                                                                                -. y 7-l,;yn.            v s:.f        &.Mr>C.n.:---/..,                        n                              y,                                      vw.1        .>
                                                                                                                                                                                                          ,-                                                          .mz 66                                                                                                                .m' p,                                                              9pg:;pm
                    +                                                                                                                                                                                                      >                                                                          ,      > .; .1                                z. . p v ,- ,. ;@, , , .
                                            ..x smz                                                ,
s    x ,.              .                                                                                                                                                    ,
                                                                                                                                                                                                                                                                                                                                        ,. s . "                            r .,
b; v                                                #            '
6h j                                  . . . ,
                                                                                                                                                                  / M*
                                                                                                                                                                                                                                                                      ''l-~*              ~            #
                                                                                                                                                                                                                                                                                                                      ;                  : s.
* f q,                  l'[$ib .                                              *
                                                                                                                                                                                                                                                                                                                  .?                , $
J> .l-                ,
* 3
                                                                                } ,K '[+ 'l:,Qy
                                                                                            ,>>  :- e -.                            p ; _;, , . .                        .
s
                                                                                                                                                                                                          - Q j        ...a,
                                                                                                                                                                                                                                                'M:'
                                                                                                                                                                                                                                                                                    +                s11        ' s . ,:. :              j :j
                                                      ' *                                              '                            :....f ' : ,                                                    >
                                                                                                                                                                                                                                                                                                          "r                        . . ~. .9 c
                                                  +
x [jf.4ij >.4[' '                                                                                                              > %                                                                                                                                                    '.%,
                                                                                                                                                  ..'Jh  * - , 4*      y "~ se:
4.rGig
                                                                                                                                                                                                                                                                                                                    ^yf ( n @:l?.>' M                        . .MQ
                                                                                                                                                                                            ~ ''                                                                                                    '
            . ;g ,,'. .. w
              'cn. r.;Qgc.) 6 i                                        Q -{>;
s ;g -                                                      :,j:<                        iR:  1. n .- .
                                                                                                                                                                                                                                                                                                ; .y_9                                            ,
V 'k m
:N ,                                ,- ,
                                                                                                                                                        *3 %
3 " A:                                      $ ';"'                        ' 'i                                    V
                                                                                                                                                                                                                                                                                                  -x'^                                  5
                                                                                                                                                                                                                                                                                                                                                        , . . . ( 7,'#c/p%    e i
    ,9                < <j s                  m: ;                                                                                        i                -
                                                                                                                                                              ?fi j                          v
                                                                                                                                                                                                                        ,        <          ~ s s ,        s
                                                                                                                                                                                                                                                                                        ,q.3                                s                .
7,g. - j i                                  .
                                        > ~          >F          >                                          c , . .s Q.) 4                <      s
                                                                                                                                                                                                                      <          , ,                            .i                  QQ ;0 '                                          f$' f - f            f'?)'I' '                            lfi
              ,,>    ?.;-).l>y. l    :                        , . -y'h.,  ;                      ,.,                                    s Y
v
                                                                                                                                                                                                                                      ?        x v
f:'d.
                                                                                                                                                                                                                                                                                                                            - lN.'&;;t u -              l              :
                                                                                                                                                                                                                                                                                                                                                                  -v .
                                                                                                                                                                                                                                                                                                                                                                                      -.a,.,.,4,s.
          ,. <            >      .M
                                                                                                                                                                                                                                                                                                            , . . . . ~ +
mv.. .e y= w :%.-                                  3
                                                                                                                                                                                                                                                                                                                                      . - Y. gM.I
                                                                                                                                                                                                                                                                                              .x k ks--+'.O                w:../
                              *p~ y, s .
                                                  <2 i.
UN.[
                                                                                  - > ..rp
                                                                                                                                                                                                                                                                      .r d
h-4v:s  - n.      ~ ,
i''?                  s                            ryg
                                                                                                                                                                                                                                                                                                                                                            . A m, ..
g      .*
h-  h
                                                                                                                                                                                                                                                                                                                                                                                        .lf@y r
f.pyy                                  ['                                s,                                      ^O TN.                        O.M                            V:                  s    @s .,    G            ,      Q.
        ' q,                N                                                                                                                                                                                                        pg                                                  i x~nA44-        .M            w@ .2.,:. d3,sgMr;.
                                                                                                                                    ~
                                                , :m,. ,,g w i;r.p
                                                                                                                                                                                                                                                                      ^
4 w, , .pw.~w.              : .
p ys u.,, s.<
                                                                                                                                        . s a. . ._                  .s g3, x            ;
                                                                                                                                                                                                                                                                                                              >j    ;.
                                                                                                                                                                                                                                                                                                                            ,is, :. w w- v;                    s y
                                                                                                                                                                                                                                                                                                                                                                              %y:g:        a g pg 5." ~ s e(.,q, x t ,. o ., j ;'
                                                                                                                                                                                                                                                                                          ,v y<                          ,                                      c.                        (                                                                                                                                    . 9. s. -g.                :  :
se                    v eh                          <
ss                  cys                                                                                                                                    -
                                                                                                                                                                                                                                                                        ,      34                            yv/ . <                          4,      m s
        's                          'e              *4  . .. iS:J
                                                                >                                                                              b.                                                                                                                    ?'                                        n.            $$ ,        <                          ?                QN RsR.    .,o
        , i v-
:/ g ; <
                                                +
                                                        %., . W                b>g-                4 w.....
y .-
                                                                                                                                                          #.4 6                                                                          'f%  , , .
u,. 3::.j..
                                                                                                                                                                                                                                                                            - :pf v
                                                                                                                                                                                                                                                                            .) ' +
WWW 5
                                                                                                                                                                                                                                                                                                                                ...<c
                                                                                                                                                                                                                                                                                                                              *q u W      I4.,v 4 3,5.:. A,s y:
                                                                                                                                                                                                                                                                                                                                                              .e
                                                                                                                                                                                                                                                                                                                                                                              .6^C,,s      g
                                                                                                                                                                                                                                                                                                                                                                                  ,; y y 3 ;.
a.
                                                                                      .v;..-.
                                                                                                                                                                                                                                                                                                                                                      "N
                                                                                                                                                                                                                                                                                                                                                      -N (*@. c. ,, .[ > ' '{
si' M        . :13 J'                                                                                        .: W."              VD:''''9'"                      %.t.':y. :q          .F.c.~.,
:                                                    ^J/ v):
e                x                                                                    '' .;?f'?;6          . s.k.g        : -, v sx                                                                                                                                                        -n                                    QV ^@;E @-t Q'd.4.h ?
      ,              ,                      *g        _.;<l y. ,-                      ' '~
N,:  v: 9^                g'f7t. , ss                                                                                ~
    ,      y    5.  > : 3 ,                        _.;g t              ,
k                          ~ ~c::                                                      ,  yte
                                                                                                                                                                                                                                                                                                                          ; r :( a ,.. ,p      gy;;      >  :.~g,9    s
                                                                                                                                                                                                                                                                                                                                                                                    -< f vg -,' + .,
n,
      ?.
y {,, f      ,.m              ' r-p4s Q9+
* T.l >
gM)5                                                                              > .<                                    ., , . " .
bk
                                                                                                                                                                                                                                                                                                                              . t . 3. L w,        #
                                                                                                                                                                                                                                                                                                                                                                              ' : :J h.
we- -vs<
                                        .            5                  s                                            4 3                                      %: 4 x :-              'yg
                                                                                                                                                                                                                                                                                                            $r M%:S {gg'3W:                                                            >
s                                              :                                                                                                                                                                                                , y y< %ab:,:; . m b;.b %,M N:. :^ t.g ^s l::.-@@c_
s..
s,.-.,-'
                                                                    .'y:g 4,+
_.g
                                                                    .s,.
                                                                                . aR :          ,
x
                                                                                                                                                                                                                                                    , . ~ . ,                                        >          % :,.. .y>            .e..;.,.>..w.xy m
: n.            8 w4
_,s v
[h :h,h>N diR;. s                                                                    3                                                                                            ;:q?
* 8 ^;k l[l%                                5 U.l !$                                                  &                                                                              _ff                          ,l<>                                          _
0<. !>
                                                                                                                                                              #                                                                                                                '        s MB r. t.                                          n.
3' ' . . . .
                                                                                                                                                                                                                                                ,                                                . , 2. . '
                                                                                                    ..*                                                      .M                                                                            + ::qjp-                                          N, yg NJ
                                                                                                                                                                                                                                                        .g .f
                                                                                                                                                                                                                                                                            >p.s    N;iw. -u,
                                                                                                                                                                                                                                            ;s99.x.9 w            d.  -
V.r s                                                                              ~.3 i
CAR 35 B SUPPLY AND RETURN PIPING
 
CEDM Cooler Units m Physical Arrangement m Performance Characteristics O
e m-
 
                                                            .i
;      ComputerModeling m RBCCW System
                                                              ~
Components
* CA.R Cooler Units
* CEDM Cooler Units 1
l l
[
i
 
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                                                ,      .I.                          .I.            S 6
I
                                        'n
                                            .$.                    .I.        I h      .I.  . h 3      0            $            C                  E    R g    *      *            *
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                                    -O 2
h  i
                                                    .e.
5                          .u.                      .a.          2
                                                                                                          .        R E
O g
E      LI        'I E          3      E    E          5 I'  I      I            I          I        -
I    I      I        I h
3
                                                                                                        .e.
K I
1.
a
                                                                "                    I                a s-gu
 
CEDM COOLER UNIT MODEL 1
l E
b          6M
* SINGLE TUSE 0 h y
g Er  h                  4 Or 630 625                409 h
COMBINED 622 623      TUBE    627 629 621          . PPER COIL 411 412 l
508                  LOWER CO!L 634    635        410 633                    '
COMBINED TUBE        636 632 631 512 l
l l
l                            CEDM X-348
 
Analysis a Sensitivity Study                i m Scenarios
* Restart Time
* Surge Tank Pressurization
* Non-condensibles in Water
* Orifice in Surge Line -
 
Results ofAnalysis m Observation
* Formation of Voids
* Collapse of Voids-
* Acceleration & Deceleration of Fluid
* Pressure Responses
* Transient Piping Loads              .
 
MILLSTONE UNIT 2 LOCA/ LOOP STUDY Header B - Sequence 2 Pump Restart at 26 sec 600.0                    .                  ,              ,              ,
i mflowj-201930000 l                            - mflowj-213010000 400.0 2            '
        'E e
200.0 Q
cc                                                                      ; 4 E                                                DOWNSTREAM N                                      '
O.0 -
UPSTREAM
          -200.0 0.0              10.0              20,0            30.0            40.0        50.0 Time (s)
FLOW RATES UPSTREAM & DL /NSTREAM OF CAR COOLER UNIT
 
MILLSTONE UNIT 2 LOCA/ LOOP PTUDY Header B - Sequence 2 Pump Restart at 26 sec 300.0                ,                ,                ,                  ,
                                - p-727010000 p-727040000                                                                :
p-727080000 p-727140000 200.0  -
      'i5                                                                        l t
t
      !                                                                                                        i e                                                                                                        i
      '                                                                                  l 100.0  -
l p w~                                                                            th
                                                                              ~
p A      - --                          .
I 0.0        -                -                  -                -      '
O.0          10.0              20.0              30.0            40.0                50.0 Time (s)                                              .
PRESSURE IN CAR UNIT TUBES
 
MILLSTONE UNIT 2 LOCA/ LOOP STUDY Header B - Sequence 2 Pump Restart at 26 sec 300.0                .                      .              .              .          -
p-727010000 p-727040000                              i p-727080000 p-727140000 200.0    -
7                '
                    'Di 2
8 m
b                          i n.
l 100.0  -
r'    gf 'Ny
                              "                  ,        ;          1 1
{.
l      1
                                            '                      '              '                                            ~
0.0      -                  -                              -        '        '
36.0          38.0                    40.0          42.0        44.0                46.0 Time (s)
PRESSURE. A CAR UNIT TUBES 3
 
0
                          ~      -          -        -                    ~                0 5
Y D
l                            '
U 1
0 S
T    c e
                                              ,                                      ' 0 4      R S                                                                l .
E
                                                                  \
s 6
D P  2                                                                                        A E
Ot  a                                                                                        H Ot L  t r
a T
E s                                                                                        L
      /    e                                                    '
0 A                                                                                            N R    .
0      I C
p                                                                                3      T m                                                                                        I O  P u                                                                                  )
s N
U L                                                                                        (
2
_              e  R 2    e c
mA C
i T    n                                                                                  T I    e                                                                                        N u
N    q e
0 0
I E
U  S                                                                                2      R U
E  B S
N    r      000 000                                                                  -
S E
Od  e      000 000                                                                              R P
T    a e      21 4 000 S  H      255                                                                      0 C'
L        .
000                                                                      0 777 L
I ppp
                      -      -                                                              1 M                                                              i                -
                      -                                ]' \
                                          -        -b 0
0        0      0        0      0 0        0
: 0. 0 0        0      0                                    0 0        5      0        5      0                        "
3        2      2        1      1
                                    %*i0 B
23Ee o.
 
e MILLSTONE UNIT 2 LOCA/ LOOP STUDY Header B - Sequence 2 Pump Restart at 26 sec I
140.0      -
p-121010000 120.0      -
100.0    -
2 Tn O      80.0    -
e 1
60.0  -
p=        y-                e-
                                                                            }
l w            -  -
40.0  -                                                        .
20.0 -
0.0 l              0.0              10.0                20.0          30.0                ,40.0            50.0 Time (s)
PRESSURE AT SURGE L. 2 CONNECTION TO SYSTEM
 
MILLSTONE UNIT 2 LOCA/ LOOP STUDY-Header B - Sequence 2 Pump Restart at 26 sec 400.0            ,                ,            .              .
mflowj-125020000 j    200.0  -                                                                        -
I e
2    0.0  -
                '=                                                                -
j E
e
    -200.0 -
4
'  -400.0    -      '      -                      '      -      '  -
O.0      10.0            20.0          30.0          40.0            50.0 Time (s)
MASS FLOW IN SURGE LINE
 
Conclusions i          m Voiding occurs in the system.
l j          m During pump restart voids collapse with l
associated fluid acceleration and
!              subsequent deceleration.
m Pressure surges during fluid deceleration { void collapse}.    .
m Pressure transients and the piping loads are not excessive.      .
 
i I
g Calvert Cliffs Nuclear Power Plant      ,
Calvert Cliffs Nuclear Power Plant Generic Letter 96-06 Activities i
f i
J. Todd Conner                                          !
Mechanical Design Engineering                                i (410) 495-6961                                          l i
l 12/2/97                                                      said.1
 
w Calvert Cliffs Nuclear Power Plant Description of S_ervice Water System
            -  Two independent Safety Trains
            -  Non-Safety Loads are Cross Connected and isolated under DBE
            -  Closed loop system
            -  System is vented to atmosphere @ head tanks.        .
i
          ~
slide 2 12/2/97
 
  \lli:ll          ,
~
                                                                                        ~
t
                                                  -                                    4 7
n l
a P                                                                                  3 r
e                                                                                s m'.
                                                                                      ,        i w                                                                    (,    -
i s
u-
                                ._                                      l o                                                              H e        P P                e                                        , 2 W
R S
r          s  r e    r s
{t a          n iw e        dn  t
                                                                ~
a l            e H
c                        2,      (
u              w          t r l
e C      g' m".
H N
                                                  ;i t
i 1
* s                                3  i f
f i
l h  r
                                                          )
                                                            -M!  '
                                                                  'r ,
C                                        _
1    i t
r                ;
                                            , A C    _              -
e            "
                                            ,C l
v  .
a                    __
C
                'd~
l                        .:**!:
f    .
1 y                  C .
                                            , A c
A ,                , ,
L y              g y
_                7 9
                                                                                                  /
2
                                                                                                  /
2
_    1 k
L
 
t Calvert Cliffs Nuclear Power Plant l
!                                important CAC Facts
                -  Each train has two coolers, one at 69' and one at                        .
45' Elevations Each CAC is flow controlled on the inlet Design accident flow is approx.. 2000 gpm Coolers are American Air Filter l
Tubes are brazed vice roIIed CAC's are required by Tech. Spec.'s
                                                                                          . i 12/2/97                                                  said 4 i
f
 
O Calvert Cliffs Nuclear Power Plant c_., -- c              #
GOTHtC Model to                              1
                                                    >    HSTA Modet for      --
                                                                                      *h        g 4 g g, and pressures
* M pressures approt 1180 psia 11/4/97                                                            12/14/97  -
[Sargert and Luney] fiows CAC mR BM [3gL py, ,m,]
si is sec.                . , ,,ct g Performs Hydraune                          *                        -
ana8yses for Watar
                        )
12/05/97 f~        ifo3sy                                          1/17G7                    INI Generate ME10                  Perform Trarsrent                  Compare Suport                                        Prepare and Sutumi structural Model and        ,      W tofind                  ,    Loads and ppe        _,                          ,    Response to me develop support              Ppng Loads and                    stresses aganst                                              NRC w.g    mm.                Support Reactions                  operatzmy entena                                          By ot/2ss7 1/10S7            t/10/97                      1/17/97                1/24/97 c            ,        r                m        ,                  m          r            m            e-
  ,,= g,,,,
                    -,    Tu Mre.r
                        < Two concepts >
n"*gaod**" -.
g=
J-
                                                                                                    -fk      W>ched b/        i            3,,se7 r                          r                ,
                                                                                                                                            ,  nar. . -
engmeenng packM                Modificamon BGE's Resolution Approach                                                                                  t              J          t                >
k    12/2/97
                                                                                                                                ~
stie s
 
Calvert Cliffs Nuclear Power Plant Operability Results The lower CAC's (45') will not boil prior to pump restart The upper CAC's (69') will boil in 12 - 15 seconds &
Generate ~30 ft^3 of steam (completely void CAC)
The differential impact velocity will be ~13 ft/s, or pressure pulse of 1180 psi (vs ~25 psi)
        - Assumed that the vapor pressure goes to zero at pump start All piping and supports met operability limits.
CAC coils were found sufficiently rugged (Brazed Tubes).
12/2/97                                                              slide s
 
Calvert Cliffs Nuclear Power Plan The Inputs to the Decision to Modify....
                    -  At the October 1996 Workshop
                        - PGE presented their experience with a similar system, w/
similar results ....They implemented a modification to eliminate the event
                        - The NRC provided strong guidance with regard to the treatment of this GL, and that it was ' extremely safety significance'
                    -  NEl Correspondence                              -
                        - Relayed the NRC's emphasis that this was a significant issue
                        - Indicated that Units in Outages may be able to restart under            7
                  '        91-18, but, NRC expects everyone to follow 91-18 [to the letter]
sm7 12/2/97 L_.._____________
 
I
~
Calvert Cliffs Nuclear Power Plant l              The inputs to the Decision to Modify....(Cont)                                                                      '
Industry communications
            - Units going into outages felt pressure to implement corrective actions on ASME Piping and Pipe supports per 91-18 based on recent industry experience CCNPP was entering an outage in 8 weeks
            - Margin on operability was small .... consistent with high safety significance....CCNPP results similar to other plants'                                    .
            - Cost of 1 week start-up delay exceeded the anticipated cost to modify.
(      - The modification was found to be constructable within the time constraints.                                                                          -
                                                                                                                    ,l 12/2/97                                                                                              slide a
 
f                                                                                            i Calvert Cliffs Nuclear Power Plant I
Highlights of the Modification Approach 2 weeks were spent studying options and pre-scoping                                      ;
1 week spent detailed scope and cost estimate
^
A standalone engineering office created with 20+ engineers                              !
2 weeks to develop the 50.59/ USQ submittal ( in parallel with scoping)                                                                                !
4 weeks for overall design, and identify available parts Created an independent review team to find design flaws.
Modification was planned and completery constructed with no '
impact on the outage critical path.
1                                                                                              t 1
3    !
12/2/97                                                        slide s                      l i
 
Calvert Cliffs Nuclear Power Plant USQ issues                                                      ,
Determined to be a USQ because of dependence on a new design feature (i.e., Nitrogen)
Received first request for additional information in April Contrary to 91-18 and strong NRC (NRR) guidance provided at                            ;
Oct. meeting, the region indicated that there was insufficient time to adequately review the USQ request, and suggested that                                !
CCNPP should consider not implementing during the Spring                                !
1997 Outage.                                                                            l BGE chose to perform all work necessary short ofpressurization, in case we decided or needed to implement during a forced outage or at power. (Over 1 million dollars in expediting fees)                                      .
12/2/97                                                            said 10
 
;  , M                            caivert Cli#s Nuclear Power Plant As Result of NRC relaxing the previously stated position....
Resources could be directed to the other unit to avoid " Rushed" modification Able to study event more completely Able to cap costs, and act less reactively.
12/2/97                                              Siide 11
                          .                                              j
 
Calvert Cliffs Nuclear Power Plant                                .
Results of More Complete Review                                                                    ,
Identified an intuitive / conceptual inconsistency in A/E's void collapse approach
        - The steam pocket will be surrounded by near saturated water. How can the void collapse without a sink for latent heat of vaporization?
        - Vapor pressure cannot go to zero psi as assumed Teamed with Creare to analytically model the void collapse Obtained third party review of new approach (Dr. Peter .
Griffith formerly of MIT)
Developed revised forcing functions and re-evaluated piping and equipment                                                                            .
\                                                                                                          ~
12/2/97                                                          siin 12 i
 
i i                                                                                            i
'!                                                                                      -l Calvert Cliffs Nuclear Power Plant I
r l
l l
l
^
tumted Water @ 2g4y i
Steam @ 264F l
l l
N Saturated Water @ 264F
                                                /
!      12/2/97                                                slide is i
 
M                                                                        caivert cliffs Nuciear Power Plan Results summary 1
        -                          Pressure will be less than 300 psi                                                ,
Pulse shape will be triangular
        -                          Pulse Duration will be approx.100 ms Piping models and support calc.s are still being finalized.                                            .
Draft piping results demonstrate full compliance without modifications
      ~
12/2/97                                                                                            slide 14 i
 
Calvert Cliffs Nuclear Power Plant-
_ Current BGE actions All modifications put on hold until finalization of                          l re-analysis Expected outcome will be a revised submittal of the Jan. 28,1997 response
                    - CCNPP will boil in the higher elevation CAC's
                    - A small water hammer may occur which will is bounded by a 300 psi pressure pulse
                    - BGE has incorporated water hammer loads into design basis, and is in full compliance with licensing basis
                    - No modifications are required                              .
J 12/2/97                                                                      siin is
 
Calvert Cliffs Nuclear Power Plant        ,
_ Causal Factors
;                                                            BGE made a large financial and resource investment, with little to show....WHY?
NRC Related factors                                                              l Fast Track" issuance of the generic letter contributed to the perceived safety significance of the issue                                    1
                                                        - Fast Track" issuance of the generic letter bypassed public comment process
                                                        - Generic Letter 91-18 restrictions of operability time limit for
* ASME piping and supports contributed to reactionary approach.
Requires resolution at next refueling outage.                      .
                                                                                                                                        ~
12/2/97                                                                slide is
_ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ -            - - -                                                        4
 
Calvert C!iffs Nuclear Power Plant l
Causal Factors (cont)
    -  BGE Related Factors:
        - Timing of CCNPP Unit 2 outage was such that decisions were required quickly.
        - The tight schedule limited study of the problem, allowing the inflated safety significance to be perpetuated
        - Economically, the high risk of start-up delays and related costs, far exceeded cost of the modification.
Other Factors:
        - Results were consistent with other plants of similar design    ,
        - Other plants had either reacted on their own initiative, or with regulatory pressure to correct in an expedited manner 12/2/97                                                          slide 17
 
Calvert Cliffs Nuclear Power Plant What Can Be Learned
    -  NRC needs to ensure adequate public comment and review is allowed.
    -  GL 91-18 Attachment 1 guidance for ASME Piping and Pipe supports should not be limited to the next outage, but instead be consistent with the rest of the Generic Letter.
        - Revision 1 is a step in the right direction, more is needed, Still need to delete ASME restrictions.
Need more consistency between NRR and Regions                                                                      ,
i i
i \
12/2/9'/                                                                                                          sude is
 
Calvert Cliffs Nuclear Power Plant What Can Be Learned (Cont.'d) l
                -  Licensees need to be able to open a dialog early on.
                        - Needs to be two-way communication, not a directive from the                  l
                                                ~
staff.                                                                l
                      - Must be careful not to paint all plants with the same brush.              ,
Industry Groups need to be more active, both technically and in regulatory space
                              - Serve as a conduit to related experiences, and communicate with the Staff                        ,
                    - Help moderate industry response to ensure that one drastic reaction does not force drastic actions by the NRC to be applied to all.
4 slide is 12/2/97 e_-_-_______.        _ _ _ _        _ -  _.
 
Calvert Cliffs Nuclear Power Plant What Can Be Learned i
L Do not ' lock-in' on one approach...Always question where you are regardless of path to get there.                                                                                            .
                                          - Current set of constraints may result ... a different response
:                                            which can be more cost effective, or enhance safety.
1
                                          - Challenge all assumptions, especially sensitive ones.
Independant Assessment of" Fly-up" modifcations can prevent many conceptual and detail problems later                                            ,                              ,
12/2/97                                                                                                                    siin 20 t
                                                                                                                                                                                }
 
i i
l THERMAL OVERPRESSURIZATION                                                      -  '
I i
* BACKGROUND                                                                              '
4 e LICENSEE RESPONSES i
l
* GSI-150                                                                                -
i
* ASME CODE CRITERIA e  CONCLUSIONS I
LEAD REVIEW: MECHANICAL ENGINEERING BRANCH RICHARD WESSMAN i                KAMAL MANOLY i                GARY HAMMER I
JOHN FAIR I,
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l THERMAL OVERPRESSURE - BACKGROUND                                                                  -l i
e  Thermal overpressure is caused by heating of trapped fluid in piping during a scenario such as a postulated LOCA or 1    MS line break accident.
i i
e  Heating of fluid in piping runs containing cold fluid by 200 4
degrees or more is possible during these accident
;  .scenanos. This heating can cause high pressure in isolated l    sections of piping due to the suppressed expansion of the i    trapped fluid.
i l
 
i l          THERMAL OVERPRESSURE - BACKGROUND                                      .
!  e The piping design codes have a specific paragraph in the                      ,
{    design considerations that addresses this issue.
i l
ASME Code NC-3621.2 Fluid Expansion Effects. When l      the expansion of a fluid may increase the pressure, the i
piping system shall be designed to withstand the l      increased pressure or provision shall be made to relieve the excess pressure.
;  e There have been several discussions regarding this criteria
,    during the recent ASME Code meetings.
The ASME has issued ASME Code Interpretations.
l A proposed ASME Code Case regarding NC/ND-3621.2 l      was initiated.
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THERMAL OVERPRESSURE - RELATION TO GSI-150 4
e GSI-150 addresses the overpressurization of containment penetrations.
e In 1991, the staff ruc.ommended in NUREG 0933 that the issue be dropped based on an assessmertt of the risk associated with a containment bypass event.
l e The GSI-150 resolution using the risk assessment did not conside~r the compliance aspects of the issue.
e The GSI-150 risk assessment used a simplified assumption of uniform hoop strain in the pipe wall to estimate failure i                      probabilities.
 
i THERMAL OVERPRESSURE - RELATION TO GSI-150 i
e    The scope of GSI-150 was limited to the piping between containment isolation valves whereas the scope of GL 96-06 is broader and involved system functionality.
l e    The resolution of GL 96-06 is not tied to the closure of GSI-150.
                                                                  =t
 
i THERMAL OVERPRESSURE - RECENT ASME CODE INQUIRIES
* ASME Code inquiry NI 97-008 4
Inquiry indicates that if an evaluation demonstrates that the total strain due to the event will not rupture the pipe and the stress exhibits the self limiting characteristics of a secondary stress, then the stress can be classified as a secondary stress for Code evaluations.
The inquiry indicates an evaluation of the loading condition is required.
The evaluation will require the development of an acceptance criteria to implement.
                                                                                    . _ _ - _ . __l
 
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I N            h    e n            t ede      l l      a S              y          t t e        e    ai r      ii S              r            s d et e
sh        i cuq          wre E            i u                    o        d              ni r t
R              q          t a dndi n            n e  r      oc P              n            c aFd            i            i s      a e t
R            I            i d                    yi E              e                          e    r u c V            d              nd ei xd          i u n        l      n t d u            qi o        aa i
O              o            yanl                            vt C              r ue        c    nt i        ep L                          i n  i d
A            E              ul qapi                en          e e  c M            M              n vpe eAb h
Tc o      h c Ta R          S            I E          A                -                  -            -
H T
* 1          lllL            l
 
i THERMAL OVERPRESSURE - PROPOSED CODE CASE 1
e The ASME is pursuing a proposed Code Case relating to NC/ND-3621.2 to address the thermal overpressure issue.
The current proposal contains a recommended strain limit for thermal overpressure events associated with ASME                j Code Level C and D loading conditions.
e  The NRC staff has voted negative on this proposal.
An adequate technical basis in support of the proposed strain limit was not presented at the Code meetings.
 
THERMAL OVERPRESSURE - ASME APPENDIX F CRITERIA e  Appendix F of the ASME Code contains criteria'for inelastic analysis of components.
The criteria is stress based and applies to primary type loads.
        -  The use of the inelastic criteria will require development of the material stress-strain curve.
it will be necessary to include all identified loads in the inelastic evaluation.
4>
  --                                                                      a
 
THERMAL OVERPRESSURE - ASME APPENDIX F CRITERIA e  The use of Appendix F criteria will likely require an FSAR revision and staff approval for most facilities.
Appendix F is referenced only for Class 1 components in most ASME Code Editions.
      .Most FSARs contain explicit references to load combinations and Code acceptance criteria.
* The staff is receptive to licensees' requests to use the Appendix F criteria for long term resolution of the thermal overpressure issue.
 
l        THERMAL OVERPRESSURE - CONCLUSIONS e The NRC Staff still considers the potential for thermal                            1 overpressurization of piping systems penetrating the containment to be compliance backfit issue under the provisions of 10 CFR 50.109.
l l e The NRC Staff still considers the potential for thermal l  overpressurization of piping systems that penetrate the l  containment a safety concern that should be addressed by l  licensees.
1 t
                                                                                    -i
                                                                                    .i
 
l EPRINPG                                            -
GL 96-06 Containment Penetration Piping Overpressurization Avtar Singh Phone: 650-855-2384 Fax: 650-855-1026 e-mail: AVSINGH@epri.com NRC/NEl Workshop ori GL 96-06 Gaithersburg, MD December 4,1997 i
SARA 1085V/AS/tli 8/6/97 1 l
 
EPRINPG                                                          i Overpressurization of isolated Piping Sections Under Postulated LOCA Overall Objectives of EPRI Project To develop a credible technical approach to resolve the GL ~96-06 containment penetration overpressurization issue and minimize unwarranted hardware modifications To develop simplified elastic-plastic methods that can be applied by utilities to evaluate these    -
loading conditions without the need for detailed finite element analysis SARA
                                                                                        .l 1085V/AS/tfl8/6/97 2
 
EPRINPG                                                                              .
Overpressurization of isolated Piping Sections Under Postulated LOCA EPRI Project Two-phase project
                                    -    Phase 1 focuses.on straight pipe tests and analysis to demonstrate margin to uurst
                                            - Under bounding LOCA conditions
                                            - To support ASME Section 111 proposed strain criteria currently under review by ASME and NRC
                                  -    Pending Phase 1 results,. Phase 2 may, as needed, include more quantitative.(typical in-plant configurations) tests and evaluations for plant-specific applications by utilities
                                                                                                ~
SARA 1085V/AS/t!!8/6/97 3
 
EPRINPG Overpressurization of Isolated Piping Sections Under Postulated LOCA Schedule
* Participation:
                            - Twelve utilities
* Project initiated.- July 1997 Utility survey completed - 7/97 Test Specification completed - 8/15/97 Testing
                            - coupon / mock up - 8/25/97
                            - 3 pipe tests - 9/20/97 .                                      '
                            - additional hydrostatic pipe bur'st tests- 11/20/97 White Paper report documenting results - 12/1/97
\                                                                                        -
SARA 1985V/AS/tli8/6/97 4
 
EPRINPG                                            -
Participating Utilities i
Centerior Commonwealth Edison Detroit Edison
* Duke Power
* l Entergy~ Operations ~
Korea Electric Power Corporation Niagra Mohawk Power Pacific Gas & Electric PECO Energy Pennsylvania Power & Light Wisconsin Electric Power Co.
Washington Public Power Supply System
* Phase I & II
        \                                                      -
SARA l
1085V/AS/tli8/6/97 5
 
EPRINPG
        /
[                                                            Status of Phase l
                      - Three straight pipe tests completed
                          - Test 1: 3" NPS Type 304 SS Schedule 40
                          - Test 2: 3" NPS Type 304 SS Sch. 40/80 Transition Weld
                          - Test 3: 8" NPS SA 106 Grade B CS Schedule 40
                      - Two additional hydrostatic tests ( 3"SS Sch. 40 and 8" CS Sch. 40) performed to assess actual margin to burst under LOCA conditions
                      . Detailed data evaluation and benchmarking of a non-linear analysis methodology underway 4
                      - A simplified non-linear approach developed to explain the data and demonstrate self-limiting nature of the loading SARA wasvasanarm e
                                                                                                                            ,j
 
EPRINPG                                                                      -
Preliminary Results of Phase l Test results of the overpressurization and hydrostatic burst tests for SS and CS pipes demonstrate a significant margin to burst under expected LOCA conditions R.esults will be documented in a White Paper report that ASME Code committee can use for establishment of a strain criteria SARA wesvms=8    77
 
EPRI Generic Letter 96-06 issue on Overpressurization of Isolated Piping Sections inside Containment PHASE 1 TESTPROGRAM S. R. Gosselin, P.E.
Electric Power Research Institute Charlotte, NC OFF: (704) 547-6067 FAX: (704) 547-6035 email: sgosseli@lawson.epri.com December 4,1997 i
i
 
EPRI Topics
* Phase 1 Test Program
* Utility Survey                        ,
* Test Specimens
* Material Properties
* Test Results Burst Tests Thermal Expansion Tests
* Analysis of Test Data
* Strain Limits
 
l      EPRI Key Points
* Design containment temperature transient conditions in operating PWR and BWR are not expected to burst isolated piping inside containment
* For energy controlled conditions strain limits should be            ,
proportional to the usable ductility of the material l
* Local peak strains must still be addressed
      =
The membrane strain criteria proposed by ASME Section 111 for energy controlled conditions, are consistent with current Appendix F safety margins established for load controlled situations
 
N e        .
EPRI GL 96-06 Test Program Phase 1
  = Examine long term operability of isolated piping sections subject to post accident temperatures inside containment Test prototypical configurations Validate non-linear models Screening criteria
* Phase 1 Scope Plant Survey Pipe Segment Testing Analysis Technical Report Review with NRC w
 
EPRI Utility Survey UTILITY            NUMBER      NUMBER            PWR  BWR SITES      PLANTS Commonwealth Edison              6                  12    6    6 Centerior                        2                    2      1    1 Detroit Edison                  1                    1      0    1 Duke                            3                    7      7    0 Entergy                          4                    5      3    2 Niagra Mohawk                    1                    2      0    2 PECO                            2                    4      0    4 PG&E                            1                    1      2    0 PP&L                            1                    2      0    2 WPPSS                            1                    1      0    1 Wisconsin Electric              1                    2      2    0 KEPRl/KEPCO                    4                    18    18 U1  0 TOTAL        27                    57    39  19 NOTE: 1. 4 of the 18 KEPCO plants are PHWR
 
EPRI Plant Survey Results l
* Pipe Sizes 3/4" - 1" NPS, Schedule 80/160 1                2" - 4" NPS, Schedule 40/120/160 6" - 12" NPS, Schedule 10/40
* Materials Small Bore Piping - TP304/TP312 SS Large Bore Piping - A106 Grade B CS All Seamless Pipe
* Insulation Small Bore Piping - None Large Bore Piping -Cal /Sil/ Fiber Glass /None a>
                                                                                                            - .. ... .      l
                                        .....i,....-... -- . ..
 
                                                                        ~
                                                                          .I EPRI TestSpecimens I
I
* 3" NPS Type 304 SS Straight Pipe - Schedule 40 3" NPS Type 304 SS Straight Pipe with Schedule 40/80 Transition Weld 8" NPS SA106 Grade B Carbon Steel - Schedule 40
 
EPRI Pipe Material Properties 3" NPS Schedule 40 S m (psi)                        19,000 Sy(psi)                          38,666 Su(psi)                          87,728 0.3 E (psi)                          2.83E+07 n                                0.43 K (psi)                          191,521 O
 
l
                                                                                                                                                                                                                        *1. l i
2 i
EPRI Stress-Strain Curve 3" NPS Schedule 40                                                                                                                                                        '
t 100
:                              r                                                      ;                                                            i i
t                                                            I h"      h                                                                                                            '
* _.
a          .,
                                            ,9_.                        ._.        , . . .
__..,y-M m        _
i t
i                                                                                                  i i
e  .-..+4.-.H...                      . . - -                      ..6--      _
                                                                                                                                                                      .                p__j
:                                                                        i J - L _..--.9 i
t                  i                        5
                .E  60                                                                                          ,                                                            j
                  -        :                        j                                                          i                                                            i e
50 - i-
                                                                                -+ - -+ i-4-~
                                                                                                                                                            - t-t +-                                                    ;
t A    40 - :;                I l-i in        '
I                                                                                                                                                                              :
a      .-
                                        ._..77__.                              _ - . - --- $        . 4 ~ .,        --..]__.      ._.j__.. 4c            ._.4-.              _._7      . ~ .      -~
:                I                                          j                                                                                                                                .t 20 - : --2
                                                                                        ~                                                                                    ,
p                                        - - + - - .                    ;                      -
10-0          .    ,.,      .        .,..,                  .i.i.o                    . o        .4      .  , .i.,              .,        .,        . n        .n    .i,.
1            1          5      5    1        1                      5                  1        1        ii                  ,,
p 0.00              0.05            0.10                  0.15              0.20              0.25          0.35              0.45                0.55              0.65                            i Strain, inlin.                                                                                                          j t
l l
6
 
EPRI Pipe Material Properties 3" NPS Schedule 80
;                                                                                                                                                                                                                                                                                                                  i l                                                                                                                                                        Material                                      SA312 TP304 Sn(psi)                                        16,400
;                                                                                                                                                        S m (psi)                                      19,000 i
Sy(psi)                                        40,239 Su(psi)                                        89,197 0.3 E (psi)                                        3.06E+04 e
 
4
                                                                                                                                                            )
                                                                                                                                                        .i EPR Stress-Strain Curve 3" NPS Schedule 80                                                                                                  -
t 100      _
90 -E--                                                      --
80 - E
                                                        /'                                                  N    -
:                                                                                                                              t c    70 =:=
1 m                                                                                                                                          '
x 60          - -    -              --            - - - -  -                  -- - -                              -
r m    50 -:-
m                                                                                                                                        l 2
40 --  .
g M    30 - :-        - -  --          --              - ---                  --      --            --
20 = 1-10 = :=                                                                                                                            i
:                                                                                                                              t 0        '''            '''        '''            '''        '''
I t
0            0.1          0.2      0.3            0.4            0.5                0.6                0.7              /!
Strain (in/in)
                                                                                                                                                          ?
 
EPRI Pipe Material Properties 8" NPS Schedule 40 i
Material              SA106 Gr B Sy(psi)                42,500 Su(psi)                69,656 0.3 E (psi)                31.125E+06 n                      0.18 K (psi)                106,000
 
~
EPRI Stress-Strain Curve f                                      8" NPS Schedule 40 80
                                ,  I                      l                !                            !                  I "l
70 -                                  41                        l    -
i-            -
l- -        -
60 -  -
                                    /
o        '
                                                      ----        --"J
                                                                          ,      !                                                I e 50 -                    -  --
                                        - + -                -
t-            --+- - i- lI -
      .M w                                                                                                                  r              i e 40 -  -
f                                  l e                                                                                              i G                    :                                                                        ;                          i is 30 -                                                                                        !
U)                                                                                                                        i l
20 - - - - - - - - -            -- -  ---
t-- - --
                                                                                                -- - - t--+              7- ---- - r- -
i                                                                                            !      !
I                          !
10 -            i p- --+;                                                                          L-
                                                                                  ---                          c-        l I      !
O-                                                  1 0.000 0.050 0.100 0.150 0.200 0.250 0.300 0.350 0.400 Strain (in/in)
 
I
~
!' EPRI I
!                            BurstTests                        '
i
.i l          Stainless Steel                  Carbon Steel I
* 3" NPS Schedule 40 '
* 8" NPS Schedule 40 Calculated Burst Pmssure
* CalculatedBurst Pressure i        Harvey - 9,627 psi              Harvey - 5,412 psi l        Roark- 9,150 psi                Roark - 5,293 psi j        Cooper- 10,435 psi              Cooper - 5,650 psi 1
* Burst Test
* Burst Test
[        Pressure - 9,220 psi            Pressure - 6,526 pei
!        Fracture E - 36.3%              Fracture E - 8.7%
j i
i i                                                                !
l
;                                                              ~.
!                                                                i
 
L. -.u-s-m-an--J.w ._ea    __s._ea ..m4-          ma        s, --.wAa:. .4LJhe_ ..e-M__m4_            .h_aan_-4m2.,,ma-Ssm.,%4A_._mya                                                ,A-.-4wm_%A44                  e_W_#+Ja_A.m__                Ju_d %,4.- u w, e      e                                                                                                                                                                                                                                            1 l
l t
i I
i d
l                                                                                                                    l                              ;
:                                                                                            _ _        2 I
_A                  _ ,.                        _.
7
                                                                                                                                                    .      y ..--    _,                      . . . _ . .            .. .
O 4
M' d-    , - - - + -                  e  m.--.rm              ,
                                                                                                                                                                              . , , , , , , q i
!                                                                        O                                        I                                                    !
o                . . 8 T                                                                                                                        ,                      4 y _g                                  _.
_.+..._._..
i
_..._j_._ _ a _
,                                                        @3                                                                                                              i                                  i                      i W                                                                                                                                                                8 b]
i d                                              '
g                                                                                              i                                                            h
\
* O                                    - -+---                                                    -
                                                                                                                                                              - S ---- - ---                                                      9
.i 1  '                                                  7..
:                                                    03 CL              g                                                                    -
                                                                                                                                                            ).__;
i                                                    :
                                                                                                            .- .                      -.-                        ,_            .                                .. . h I
a ---.
                                                                                            ....m....m..
                                                                                                    <            m i
g .
,!                                                                                    i i i lliiRR!
Isd 'eJnsseJd amme g
l
                                            ,-. - . - - , - - -                                  .    ,, - - , - - -                            ,                    , , , -          . , .--.          -a,--                          . - - - - . , ,
 
        ,l ll      l;iI)                jiIll
                                                                                                                . G 0
3
                                                              ~
                                                                            -              -            7
                    =-
2 ilij a            r!f          , 4 2
            .i'
:I:I                            g    1 2
0                      -.
4                                                                      _              -      .
tl  e                      -
                                                                                              ' .      8 1
n su                                                                      -              -
i m
ed
* t t!            5 e
Teh                        -
1 m
ts c                                                                                                    i pI
                                                            +
2 T rS                        .
5 1
uS BPN                                                                              tf+              ;9
    "                b.                                        r            g p
6 8                                                                                              .
                                                                                                +;
3
              - ' . " ~.. : -
                                          -        ::..L*~
e * -
0 0                          0              0      0                        0            n 0
0 7            M0            0 5
0 0
4 0
0 3
Q 0
1 0
Q ew::eeeb              O.
                                          ..              3 I
R P
E ji        <1      Il
 
i                                                                !
i l  EPRI
!                    Thermal Expansion Tests                    i i                                                                I l
* Test temperatures representative of plant accident
;    conditions (300-350 oFl
* Preliminary Analysis
!        Non-Linear FEA
;
* Pressure vs Temperature                              ,
* Strain vs Temperature I    a Burst Pressure Calculation
* Thermal loading in incremental steps. Allow temperature to l    reach equilibrium state at each step.                        j
!      Maximum strain at pressure-temperature state Equilibrium Criteria                                      !
* Water and metal temperature within 3-4 oF, and
!
* Constant Pressure, and
* No radial or axial displacements i
I-
 
EPRI Thermal Expansion Test Results i
Test    Material      Size    Temperature Pressure  Hoop SNain I                                      ( *F )    ( psi .)    (%)
)  Pretest SA312 TP304 3" Sch 40        305        5600      2.40
!  Test 1  SA312 TP304 3" Sch 40        323        5700      3.00
!  Test 2  SA312 TP304 3" Sch 40/80    315      5900        5.40 i  Test 3  SA106 GR B  8" Sch 40      265      4800        2.60 l
4 G
 
    .#.. . ._ma*-~=A4--ei-.i-,2-6--4        -ee---a*4AA**E-h.-4mJAA4--        w---u
* A a +aia A**n-                      -  Ee4*dha4----+-W--.e              d *- --+*-A6-**-m*w            am_._=.._hw.-__eA&wh.4_p=.mamh-a.=e_m_-n4AmeA-_
i 8
1
)
i i
Y D
i                      !
l                                                                                                                                                                                                                                    '
ea.                                                                                  l
                                                                                                                                                                =
l
                                                                                                                                                                                                                                        . N" I
I f                        >
G
.i                                                                                                                                                                                                                                  .
e 5                                                                                    i 4
i                                          :. 8e
:                                                                                                                                                              l                                                                    -
w .
                                                                                        =..% reen- $ e        .m    ee-.,e-              - . . . . ,
q%,.__,a w                                                                                                        ,
l                                            .  .- + 1 eo                        l i
E' OQ g                                                ;
t I
                                                    >=
y      . -                    - - -.
                                                                                                                                                            ~--.+ . - _
_ -.p a
i                                            .
3                            i
                                                                                                                                                                                      ~
l 1                                                                                                                                                                                                        !
Y l'                                                                                                                                                          k                                                                    -
                                                                                                                                                                                                        )                        .
d 9                            . . .          . . , ~                    .
A                                                                                    I 1                                                                      E                                                                                                          !
2 4
8 mumm                              O                                                                                  E                    8                  S O
A 'eJnieJe d m e 1, CL LLI 3
 
i
                  -        r b                L l
                  .a        y'                                ~-l              .. g g                  t Q                  l i
i i
l        i
                                                                            =
i
                        - ~ ~ - . . . .
a n
l                                            E i
I
          , (/)          ~~~~~~~~~----J  '
O                                                ;
                                                                            ~
IE
            *g 9 W$3                                    a.--
                                                          -=4 l      A i
                          ''              +-
                                                                                -s
                  <,,                      1                        7
                    &                      l Z                                                  i h~
{
M E                                                      .i... o  ,
Q.
m
 
l EPRI                                                                                                              l Test 1 3" NPS Schedule 40 SS Straight Pipe 0.035                                                                      ,
0.03      ---    - - -
                                              ------------d-d 0.025        ---- - ---                    -
                                                              --------------4--                              --
5                                                                  '
l
        .E                                                                            ;
g    0.02      -
E h
03 0.015 --                    --      - --                        <              - - - - -
o,                                                                          .                    .
O                                                                          I                    '
j    0.01    _                  }
l                      -
0.005                                  -- 1                ,
l 0          '
50                100              150          200        250                300          350 Temperature, F L ____                          -
 
R m
i g
8 co y
I                                                .
                          ~                                              .
OS              I i        C    _ . _ . _                -__    . _ _ _ . _              _i. g 2                                                                .
N W                ,      .
k y      - . . - - . . . - . -          1                              _
3l      _ _._2              -
i g
I
        .2              ,                                                .
3 y                                                                  .
2                ! ..                              . _      _ . .i. g
        ,9              i                                                .
W                !
gg              l                                                -
g
  -        E                                                  e    e  o o
N                        .                (9 3
N 8
v-g                                      ISd 'eJnsseJd LLI
 
  ,i !  . iIi          !1l      t!i!!!!                          ;l!.                        i !l? ,l                iiI\;i',i!Il!I
.                                                                                                                                                  0 5
3 n
i o                                      .l!
                                                                                                          -            +
0 0
3 i
t                                                                                                            '
s                              -
_                                n                            -
a r                    ;, i
                                                                                                                        -                          0 r                                                                                ,<l
                                                                                                                                                . 5 2
T                          -
F S                          -
e r
2    S                  I p
0 0
t a
u r
t 0                          -
2  e p
s8                        -
e/0                        -
m e
T 4                                                                                                                          T 0
j 1                                                              5 e
i 1
l                      -                              -                                                    '
u                      -
T d            i e
0
                                                                                                                                            ; 0 h                      -                                                                                            1 c                      -
S                      -
0 3
5 6      5                      4        3                  2              1                  0 0      0                      0        0                  0              0                  0 I                        0      0                      0        0                  0              0                  0 R
d9E. c].nngx              .
t P
E                                                                              .
i :1                      ,l                l            I                            \(ll                                        !
 
EPRI Test 3 8" Schedul'e 40 CS Straight Pipe                                                                                          !
                .000                                                                                                                              1 5000    - - - - - - - - - - -    - - - -
                                                            -{ - ~ ---- -    ----j---------+-  - - - - - - - -
m                              ,
j            __
a                                              ;
6                                                                  ;
B 3000                                        1 2000                                                                      l-1000-                                                        ,
1 0
50                      100        150              200        250        300                  350 Temperature, 'F                                                              .
 
I f
4 EPRI Test 3                                                                    i 8" Schedule 40 CS Straight Pipe 1
0.040    ,
0.035 -- - - - - - - - -                ---                    -
i l
      . 0.030 --                                                                                                                                      i C            :                        i                                                                                                          I 9            -
                                            !                                              !                                                          i
    ,C 0.025 --      --
                              --                                  t C            -                                                                                                                                    1 i
t g 0.020    ,
(p)          .                                                                                                                                    :
t (2. 0.015 -I                            ---
e                  +--                --                          -
O          -
o          :
I 0.010 -i --
y-0.005 - -
;        o,gno          ..
50                    100                          150                  200              250            300                  350 Temperature, F i
e
 
EPRI Test Data Analysis Plastic Model- Uniform Thickness          l l
Ar    <                L              >
r i
t
* Straight Pipe with uniform thickness
  *  ' Plane Stress                          ;
* Ignore Elast!c Strain                  i e
 
EPRI Test Data Analysis Plastic Model- Uniform Thickness Eboop - E i  + aT                      l l
1 f
866 PrM E sp =.866                + aT
(  Kt    j Where:
n = Strain hardening exponent K = Strength Coeficient oc = Coefficient of thermal expansion
 
I                                                        '
EPRI Test Data Analysis                    '
Plastic Model - Non-Uniform Thickness -
No Volurre Expansion                                !
M    'w          >l i
x i
            <                    L                  p    ,
Volume i
Expansion 1
                            / 366PrM              1 e%g= .866                  + aT
( Kt ;              &      .
THIN                    ,
L i
i
 
                                                                                                          . . + - - - - - - - - . . . . .
i l
EPRI                                                                                                                                    .
l Test Data Analysis Results                                                                                          !
j
* Protest
* Test 2                                                                                              j P = 5600 psi                      P = 5900 psi                                                                                      i
'r T = 305 0F                        T = 315 aF                                                                                        i i
;        e      ,,, = 0.024 inJin.        w c
                                                                                = 0.054 inJin.                                              i l        ec ,icoi ,,e = 0.028 inJin.      em = 0.055 inJin.
* Test 1                        =  Test 3                                                                                              i i
i 1
P = 5700 psi                      P = 4800 psi                                                                                      ;
;        T = 323 aF                        T = 265 *F                                                                                        !
e      ,,, = 0.029 inJin.        ex                                            = 0.025 inJin.
ec ,,coi,,,, = 0.026 inJin.
L i                                        3 cc.,coi,,,, = 0.024 .inJin.                                                                      i i
i i                                                                                                                                          l
 
l                                                                                                                                    t l
i l        EPRI l
l                                      ASME Proposed Code Case Alternative Rule for Evaluating Thermal Expansion Effects Class 2 and 3 Piping and Components                                                                                :
I
* Pressure increase due fluid thermal expansion i
One time event in which Level C or D limits are applicable
* CS or SS Membrane Strain < 5%
* No limit on peak strain i
I 1
I.                                                                                                                                  -
I        __ __ _      _____ ____ __      .            . _  _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _
 
EPRI                                                                i Strain Limits
* EPRI Report NP-1921
* Strain Criteria for Energy Controlled Conditions Structural acceptance is related to energy absorption capability of the structure strain limits are proportional to the usable ductility of the material
* Plastic Instability Approach Limit membrane strain to control energy absorption Permissible energy absorption established at 70% of structure capability Consistent with Appendix F margin for load controlled conditions
* Limit peak strain the prevent local ductile tearing initiation
 
EPRI Energy Controlled Strain Criteria Membrane Strain                                                                Peak Strain 1
e            -                      -  ,
sinh 3 (1- n)
Es0.7Zn                                          E=0.74                        -                      -
er sinh      J3 3
(1-nXTF) where:    Z = Ratio of Effective Strain @ Maximum Load to Strain Hardening Exponent Er= Fracture Strain n = Strain Hardening Exponent TF = Triaxiality Factor 9
 
l l                                                                                            ~
:              EPRI r                                                                                              i l                                    Strain Limit Comparison j
* NP-1921 Energy Controlled Assessment l
* Membrane Strain l                  Z = 0.577, no , = 0.43, ne, = 0.18 i
!                  E., = 0.174
:                                                                                              i j                  E , = 0.073 i
* Peak Strain                                                                      .
i                                                                                              !
ef,, = 0.527, efe, = 0.334                                                    l TF = 1.732, n , = 0.43, n, = 0.18
;                E., = 0.206 i                E , = 0.126
* ASME Proposed Strain Limit SS and CS < 0.05
 
                                                                        .l EPRl/NPG                                                  - .
l Analysis andInterpretation of the Piping Overpressurization Tests Using Simplified i
Models and Analysis Method i
)        .
                              -Presented at the i
!                    NRC/NEl Workshop on GL 96-06 Gaithersburg, MD December 4,1997
;                                    H.T. Tang Tel: 415-855-2012; Fax: 415-855-1026
! ,                          email: htang@epri.com I
HTT 7P1
 
EPRl/NPG                                                                  .
Model and Method Straight pipe with uniform expansion
                                                  - Simple thermodynamic System p = Volume expansivity r = lsothermal compressibility GL9G4)$1NRC Workshop
                                                      ---                                                          SARA -
HTT 12/4/07 p 2  .
 
EPRl/NPG                                                            ,,
Pressure, Temperature and Volume                                    :
Relationship AP =(p/K) AT-(1/r) (A V/ V)            i i
First term - fixed volume, delta temperature Second term - fixed temperature, delta volume For temperature between: 68 oF - 320 oF or 20 oC - 160 oC p/r ~        12.4 atm / oC            .
r    ~
0.475 x 104 atm t
mm ,,me m.,                                                  OARA
* HTT 12/4/97 P.3
 
l:[      ![jj i!ll                            ,Il,. l ,          \      t!            :        I  ;I i1        !'
s                                                                            r      A s
e                l  $
r r          /        R A
n                h i
                                    =                                    A          p  r      S k                _
l D        _                              riA-    h i
c
                                      ,l zsn        2 h            -^ __              _
_                              2          o  =
T                _
                                                                          ~taL i
_              _                                      l m              _
_                              L          e r
r              _              _
r o
r no f
x i
n si/      s su
                                                                          -        n L U '  l-        _
_                              L a        p  r 2 L ___                    _
_                    r                    x e A =
l
                                                              /                        r e                                                              )r e              v r n A m2x/v d              _
A o o              _
                                                              =
i s + o      l u
M              _
n                =
A sp (r N
_              _                      i, a
e              _              _
n                          xn ip-          '_
i a
r x
e = ra/
i    v  +
P              _
                                      .'    t S                        e vt s v G
P t
h                                    p o
mA u
p A o
N
    /    i g                                  o l
o            o l
R      a                                H                      V            H P
E    t r                                  -                      -              -
S                                                                                        h p
o s
kr o
W  4 C
R P N7
                                                                                                    //
9 6/ 4 02 6 1 9T LT GH
;j,          ,                                  ;i ;:
 
                                                                                                        .l EPRl/NPG -
Analysis Results for the Uniform t
Schedule Pipes l                                      AP = (p/K) A T- (Av/v) /K                                          ,
l    .
Av/ v = K(S A T/K- AP) si, = (Av/v) / 2 i                -
3"SS919-PT - 3" Sch 40 Stainless Steel                                    j APm = 381 atm> ATm = 130 *C s,,,,, = 0. 0 2 4 (Av/v), = (12.4 x 130 - 381) x .475 x 104
                                                                                                          ]
                                                = 0.058 = 2 shc                                            i sf,c = 0. 0 2 9        m = measured c = calculated l
              ,,                                                            SARA i
 
f 1
EPRl/NPG l                                Analysis Results for the Uniform                                                  '
j                                      Schedule Pipes (cont.)
3"SS924 3" Sch 40 Stainless AP, = 388 atm; AT, = 140 o c; Su, = 0.030 l
(Av/v)c  = (12.4 x 140 - 388) x .475 x 10 4
                        ^
l                                        = 0.064 = 2 Buc Shc = 0.032 l
8"CS926-1        53" Sch 40 Carbon Steel 4
AP, = 326 atm; AT, = 108 oc; Su, = 0.026
(                            (Av/v) e = (12.4 x 108 - 326) x .475 x10 4
                                      = 0.048 = 2 Sue                                                        _
shc = 0.024                                                              .;
GL96-061NRC hkJ@                                                                                                !
HTT 12/4/97 P.6
                                                                                                                *i
 
t i
l                                EPRl/NPG                                                                                                                              -!
l                                Analysis Results for the Non-unifrom                                                                                                    ;
i                                                    Schedule Pipes Assuming conservatively that only the thinner                                                                                            ,
,                              schedule portion will deform due to temperature increase                                                                    .
l                                                                                                          .
l                            No Volume
;                              Expansion                                                                                Volume
!                                                                                                                Expansion Hoop strain / volume expansion relationship l                                - Thinner section length = 1/2 of total length                                                                                          j Av/ v = Ar/r = su                .                                                            .
:                                - Thinner section length = 1/4 of total length Av/ v = (1/2) (Ar/r) = (1/2) e u m.d_
HTT 12/4/97 P.7 SARA                                                        4
 
EPRl/NPG i
Analysis Results for the Non-uniform Schedule Pipes (cont.)                                                                                                  '
2/L l2/L Sch 40 Sch 80
* Assuming i
Assuming no                                - _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - -
uniform volume expansion                                                                                                    expansion
                        '                                                                                                                      Ar                      i l
3"SS925 equal length of 3" sch 40 i
and 3" Sch 80 with transition weld (stainless steel)
I AP, = 401 atm; AT, = 136
* C Shm = 0.054 j                                    (Av/v)c = (12.4 x 136 - 401) x .475 x 1V j                                                  = 0.061 = Eu,                                              .
l'                                                  Suc = 0.061                                                                                                      -
SARA
            $?iJ4lT&**
                                                                                                                                                                      -i
 
                                                                                                                                                    . i
;                      -- EPRl/NPG                                                                                                -
Bonding Analyss i
Total depressurization conditions                                                                                      i AP = (p/K) AT- {1/ K) (Av/v) = 0; p A T= Av / v i
For AT = 140 oC:
Maximum hoop strain for total depressurization Uniform schedule straight pipe Av/v = 0.082 = 2En ;                                              En = 0.041 D          thin                          total Av/v = 0.082 = En;                                              En = 0.082                                        ;
(
4 i
m7ggagm.e, SARA
 
              - EPRl/NPG Discussions on Configurations including Bends, Branches, etc.
Segment pipe at stiffness discontinuity Determine the percentage of total length that pipe I            will deform Derive hoop strain / volume expansion relationship Using the simplified model to perform bonding analysis If hoop strain < 5%, no further analysis is needed If hoop strain > 5%, perform more detailed anaiysis
      ==-                                                      -
 
EPRl/NPG -                                                                                  .
Discussions for Configurations including Bends, Branches, etc. (cont.)                                                          '
Examples                                                                                      .
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va                                          -'
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gosinne werione,                                                                    MM                        '
 
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                      \s        w,W,                                            pr                f,./@edp
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Generie-L,6ft6(96-06                        =e %W  a
:n Meeting        y December 4,1997
 
1 1
Wolf Creek Nuclear Operating Corporation              -
Pressurization of Isolated, Water-Filled Sections of Piping 4
Agenda
:
* Summary e  Wolf Creek License Basis l
e  Location Selection
!          =  Overpressurization Analyses
:
* Risk Evaluation i
e  Discussion t
1 l
i
 
i Summary
                                                ~
      . Susceptible Locations Selected A
e USAR allows the use of ASME Appendix F for elastic analysis
                                                  ~
      . Large break LOCA analysis complete.
      . Analysis ongoing for small break LOCA
 
Wolf Creek License Basis The ASME Boiler and Pressure Vessel code of record for Wolf Creek is the 1974 edition, up to and including Winter 1974
* Appendix F is cited in Subsection NB (Class I components) in paragraph NB-3235 and in NE-3656 of 1974 edition of the Code.
The Wolf Creek USAR states that the stress allowables of Appendix F are used for elastically analyzed code components for faulted conditions for Seismic Category 1 items other than the NSSS.
                            . The Wolf Creek USAR also gives the allowable stress for faulted conditions for Class 2 and 3 piping to be 2.4 S 3 1
 
Locations Selected -
      . 7 acations at Penetrations and Potentially Isolated Sections of Piping Fully within Containment were Reviewed
      . A total of 33 locations were identified.
      . 9 locations were eliminated:
Four locations were eliminated from consideration due to configuration.
Five locations were eliminated with operating procedure changes.
      . A total of 24 locations w                analyzed.
 
Pressurization Analyses e  Analyses methodology for remaining 24 locations:
Containment Pressure and temperature curves used for various size LOCAs.
Calculate time dependent temperature and pressure rise inside pipe section.          -
l l
Calculate stress in piping due to maximum internal l                                            pressure.
Compare stress to allowables in piping.
Calculate stresses in valves and compare to allowables.
 
Y d
:                                        Thermal Loading The overpressurization analysis considered the following breaks:
* Large Break Loss of Coolant Accident (LB-LOCA)
:. 2" , 3", 4" and 6" Small Break LOCA (SB-LOCA) 4                                                              ._
 
Containment Temperature Response to the Postulated Large Break LOCA
                                                                                                                                                  . nme            cvm...
400      --    -                  -      -    - -        -                                              .m. .          .                                                  .
350
                                                                                                                                              -- LB-LOCA 300    -              -  -          -    -
q    -  -  -    --        -    -        -    -
250    - --          -  -
4I h              -              -    -          -        -            ----        -
I-        --
e 200  -                -
                                        /                                                      'A-              -    -            --      -  -        -        -    -      -  - --          -
a                    '
                                      /                                                                \        N---
E 150                                                                                                              m ;;__                                        -
p      j                                                                                                                                                  .
100 50 1                          -  - -    -  -
0                                                                            -    -  -  -
1.0E-01 1.0E + 00 1.0E + 01 1.0E + 02 1.0E + 03 1.0E + 04 1.0E + 05 1.0E + 06 1.0E + 07 1.0E + 08 1.0E + 09 Time (sec)              .
Containment Temperature Response Following a Large Break LOCA                                                                              -
 
5 Risk Evaluation            -
                  .        The 24 piping sections were evaluated for impact on Core Damage Frequency (CDF):
None of the 24 sections perform an accident
;                                mitigation function.
Postulated failure of 22 of the 24 sections has no
:                                impact on CDF.
A failure of valves in either of the two RHR suction cooling line sections would have an
;                                msigmficant impact on CDF ( < 0.016 %).
1 k
i i
              =
i
 
Modifications Final Decisions on Modifications will be based on completion of analyses and potential mdustry-wide changes to criteria e Potential Modifications include:
One location to have an electrical logic modification to assure access to an ' existing thermal relief valve.
Potential thermal relief valve installation at one limited location.
Insulation qualification or enhancement where necessary to allow it to be credited in analysis.
 
O Containment Temperature Response to the Postulated LOCA 400    - -      -                                                                        ----                      ---                -  -
                                                                                                                                                                                                  ,,,,, ,,,,,,,,,          ,,,,,,,,r , m 2 in. SB-LOCA                  l.
350  -                      -    --                      --          --                -    -    -            --    -          -  -.                  3 in. SB-LOCA 4 in. SB-LOCA 6 in. SB-LOCA                          (
g 300-    -    -                -              -o g          ,,
                                                                                                                                                                                                      . . . . . . LB-LOCA                                1 U                                                                              '                                                                        l E 250-    -
                                                                                                        '.--.#."'(                      ,  -
s-g e
gg,                                        -
2 y    -
8 g
i f,- d,p%
E 200 7
e 4                        -
3
                                                                          ,'[ -
150                -                                                                                                                                                  --  -  -      --
100                                                                                                                  -    -          ---          -    ---          -  -          -    -
50                                                                                                                  --                              --      -~      -  -    -      -  ---
                                                                -~          -    -    --                _-                              -    -    -          -
1.0E+00 1.0E+01 1.0E+02 1.0E+ 03 1.0E404 1.0E+05 1.0E+06 1.0E+07 1.0E+08 1.0E+09 Time (sec)
Containment Temperature Response Following Various Small Break LOCAs (with LB-LOCA for Comparison) u . _ _ . _ _ _ _ _ -
 
l Preliminary Results-and Conclusion                                    ,
l
      .      Preliminary results of the analysis of the various breaks indicates that:
7 locations passed criteria with no modifications i
necessary.
15 locations passed criteria when existing insulation is credited.
L 1
2 locations require field modifications.
l l
l i
l
 
Millstone Unit 2 l
i GL 96-06 Lessons Learned Steven Unikewicz g19606-1 do:
 
Screening Criteria f
: 1. Contained fluid is not liquid
: 2. System is open in normal operation and post accident conditions 4
          '3. Local pressure reliefis provided -
: 4. Contained fluid communicates with the system / equipment / component it serves. The system / equipment / component has pressure relief
: 5. Inboard isolation is a check valve allowing fluid to expand 5
: 6. Fluid communicates with containment atmosphere directly or though check valve
: 7. Penetration is a spare
: 8. Electrical Penetration W
4 I
f
 
l Thermal Overpressurization 89 Penetrations Evaluated                                                              ,
Concerns with 9 penetrations - 12 locations Concern with 1 in-containment location Strategies Procedure fixes - 4 locations Analysis fixes - 3 locations Credit for Globe Valves lifting - 2 locations Modifications - 2 locations Open - 1 location k
\.
\
gf96Cv1.cc
 
GL 96-06                                        -
Overpressuriz3tiDn Procedure fixes Penetration #  System          Procedures    Summary              Status Penetration 49  Fire Protection OP 2201        between 2-FIRE-108  workino OP 2341 A      and 2-FIRE-120 l
Penetration 67  Purification    OP 2201        between 2-RW-21      permanent and 232              changes next RFO OP 2305        between 2-RW-22      permanent and 232              changes next RFO    i Penetration 68 Purification    OP 2201        between 2-RW-63      permanent OP 2305      -and 154              changes next RFO
 
Analysis figga Penetration #                          System  Calculation # Summary            Status Penetration 2                          CVCS    in review    between 2-CH-089    no mods - starts warm and 516 in review    between 2-CH-515    no mods - starts warm and 516 Penetration 21                        RCS/PZR  working      2, RC-001,2-RC-002 tubing, small bore pipe Sampling              2-RC-003, 2-RC-45  3/8 to 3/4", review 2-LRR-61.1          operating conditions Penetration 43                        RCP CBO  in review    between 2-CH-506,  no mods - short inside 505,198            long outside O
6 o
 
Masoneilon Globe Valve fixes Penetration #      System                  Calculation #                  Summarv            Status Penetration 14      R'NS                    working                      2-SSP-16.1          lifts at 362 #
2-SSP-16.2          sched 10S piping Penetration 35      PDT Drains              working                      2-LRR-43.1          lifts at 362 #
2-LRR-43.2          downstream locked i
                                            .s      . . . . ,    . . . . ,    .
 
\
Modif,ations Recommended l
Penetration #    System    Mod!DCR #                Summary                                        Status Penetration 10    SDC      flex wedge gate          between 2-SI-651                                evaluating gate                      and 709                                      ASME/EPRI relief?
Penetration 14    RWS/ Cont M2-97026                  Thermal relief                                      working Sump                                between 2-SSP-15A/B and 2-SSP-16.1 In-Containment    CVCS/    bypass line              between 2-CH-434                            remove intemals piping            Charging                            and 435                                      next RFO remove bypass
                                                                                                                          ~l
 
                                      ~
.                    GL 96-36 Thermal Overpressurization Turkey Point Nuclear Power Plant          ,
Florica Power & Light Jack Hamm Subhash Khurana m-
 
TURKEY POINT NUCLEAR PLANT                                                                  -
d ENGINEERING DEPARTMENT
        ~                                                    -
nees GL 96-06 REVIEW
                                                                                                                            ~
22 - Piping sections (per unit) were identified as having a potential for thermal over-pressurization under the GL 96-06 review
:10_.* Piping sections were eliminated oy analysis which indicated that the valve operator will lift (pneumatic or solenoid valves)
-  12
* Remaining piping sections require a permanent fix
* Two parailcl psths were pursued for the permanent fix (1)        Appendix F Approach                      .
(2)        Physical Modifications
* Later evaluation indicated that an additional six piping sections could be realigned to other system RVs.
* Accendix F Anoroach:
            -      Appendix F is n21 part of the PTN licensing bases
            -      Requires a licensed FSAR change
            -      Preliminary analysis was performed on 4/97
            -      This effort was abandoned due to a potential USO during the Unit 4 RFO restart.
* _ Unit. 4. Physical Modifications:
            -    Five relief valves were installed during 9/97 RFO
            -    RV discharges were routed to the waste disposal system
            -    Six piping sections were realigned to other system RVs One check valve disc was drilled
* GL 91-18, Rev.1 was issued on 10/97. This revision allows licensees to reconsider the Appendix F option. Based on this change, PTN is evaluating Unit 3.
* Unit 3 Considerations:
              -    6 remaining RV installations
              -    9 Months till next outage
'              -    Economic impact and unit similarity Page 1 of 9 i
 
TURKEY POINT NUCLEAR PLANT O
ENGINEERING DEPARTMENT h.., 6                                                    -
CURRENT PTN DESIGN BASES Design Code:              ANSI B31.1 '73 thru W.'76 Add.
Loading Combinations:
Loading            Reactor Coolant    Reactor Coolant      Piping other than Combinations            System Vessels      System Piping            Class 1 P. s S.              P, s S Normal Loads          P + P. s 1.5 S.
t                  P + P. 5 S t
a, + a, s S Normal + Design        P s s.                P, s 1.2 S Earthouake Loads      P + P. s 1.5 S,,,
t                    P + P. s 1.2 S t                    c, + a, + oa sS Normal + Maximum      P,,, s S,,,          P s1.2S Potential Earthquake  Pt + P, s 1.2(1.5 S.) P + P. s 1.2 S t                    a, + a, + aos S Loads Note that for PTN, a seismic event does not initiate a LOCA PROPOSED APPENDIX F BASES Consultant:              ALTRAN Allowable Stress: ASME Saction Ill, Appendix F (Elastic Analysis) o, s 2.4 S (hoop) o, + o, s 3.0 S or    n 2.0 S (longitudinal) y Seismic stresses will not be included in Appendix F analysis for Faulted Level D LOCA pressures Limiting DBA:            Large Break LOCA Valve Criteria:          a s Appendix F P s B16.5/B16.34 Test Pressures Valves can leak to limit system pressure Page 2 of 9 :
 
TURKEY POINT NUCLEAR PLANT                      -
ENGINEERING DEPARTMENT
        ,w/                                        .
PRESSURIZATION CASES Three cases were evaluated as a progressive approach to qualify the piping segments under the most conservative analysis.
CASE 1 Peak Containment Temoerature:
This case was considered the most conservative as the piping segment pressures were determined for the peak containment temperature.
CASE 2 DBA Larae Break LOCA Containment Profile:
A conservative time dependant heat balance was performed on the piping segments to determine the peak piping water temperature based on the transient DBA tempera *Jre profile. The corresponding pressure was determined at this peak water temperature.
CASE 3 Valve Leakaae Limitina:
This case determined the pressure required to cause valve leakage to limit the system pressure.
Page 3 of 9
 
TURKEY POINT NUCLEAR PLANT o
ENGINEERING DEPARTMENT
              'D                                                -
CASE 1 - THERMAL ANALYSIS Case 1 conservatively assumes constant containment temperature profile at the peak temperature. For PTN, the DBA Large Break LOCA peak temperature is 270*F CASE 2 - THERMAL ANALYSIS Case 2 performed a transient temperature response analysis.
Pinina seaments inside containment:
        =  Conservative analysis applied a  Consider only thermal capacitance of piping and fluid
        =  Nusselt's equation used for exterior coefficient
        =  Assume all steam environment a  Piping wall resistance considered insignificant
        =  No credit was taken for piping insulation Pioina seaments extendina outside containment:
A conservative analysis was performed to determined the maximum length for thermal attenuation based on the following:
        =  Worst case piping insulation outside containment
        =  Infinite peak temperature heat source at the penetration midpoint
        =  The piping outside of containment was modeled as a circular fin a  Considered one-dimensional conduction with radial heat !acs
        =  Applied boundary conditions for infinitely long fin
        =  Length was determined for a 90% temperature degrade
~
l Page 4 of 9
 
-                                TURKEY POINT NUCLEAR PLANT o
ENGINEERING DEPARTMENT CASES 1 & 2 - PRESSURE PROFILES
* Three pressure profiles were developed for Cases 1 & 2
* The fluid volume increase was conservatively based on only the
.        temperature rise (constant Bulk Mod.)
* Final pressure = initial pressure + pressure rise Three Methodoloaies Aoolled:
: i.        Pressurizaticn with constant pipe volume 4
ii.        Pressurization with thermally expanding pipe volume iii.      Pressurizatiori with consideration of expanding pipe volume due to thermal + pressure rise
                                                                                        ~
Page 5 of 9          I l
 
I TURKEY POINT NUCLEAR PLANT O
ENGINEERING DEPARTMENT
                *i..,e                                          -
        . CASE 3 - VALVE LIMITINC PRESSURE                                        .
No+9 that only very minor leakage of a short duration (weepage) is required to depressurize a water solid piping serment subject to thermal overpressurization Three mechanisins for valve leakage were considered:
A.      Pneumatic or Solenoid Valves:              .
Pneumatic or solenoid valves may lift if the pressure force applied below the valve plug exceeds the seating force B.      Gate Valves:
The high pressure side of the valve disc will dt;flect under the pressure load to provide a leakage path to the valve bonnet. A valve bonnet equalization ilne (if installed) can provide a bleed to the low pressure side C.      Valvo Bonnet:
The internal pressure applied to the valve bonnet can cause the bonnet studs to elastically stretch providing a leakage path to atmosphere.
The maximum system pressure was determined as the pressure required to initiate leakage where considered.
s.
Page 6 of 9
 
[I                        TURKEY POINT NUCLEAR PLANT                              -
o              o ENGINEERING DEPARTMENT PIPING ANALYSIS RESULTS Seven piping segments were evaluated 2.1d were found to satisfy Appendix F criteria for the following cases:
Seament        Case 1    Case 2      Case 3 1                      X            X 2                                    X 3                      X            X        .
4          X          X            X 5          X          X            X 6                      X            X 7                      X            X Totals            2          6            7 Note that all of the piping segments we;re found to satisfy Appendix F criteria VALVE ANALY_ SIS RESULTS
* Five valves were found to be within B16.5/B16.34 hydro-test pressures.
* All of the valves were found to satisfy Appendix F criteria
* Only three valves required leakage for qualification Page 7 of 9
 
e-TURKEY POINT NUCLEAR PLANT ENGINEERING DEPARTMENT PTN
 
==SUMMARY==
* PTN elected to make physical modifications to Unit 4 due to the potential of an USO preventing start-up under old the C 91-18 criteria
* PTN has six remaining relief valve installations for Unit 3.
The submittal of a License Amendment for Unit 3 is still under evaluation Considerations:
Cost of license submittal vs. remaining installations Dose considerations Associated maintenance and Section XI testing Timeliness of NRC licensing approval If tha licensing amendment is submitted and accepted, PTN will consider removing the Unit 4 relief valves Page 8 of 9
 
                                            /
r                                                      TURKEY POINT NUCLEAR PLANT            -
O ENGINEERING DEPARTMENT
  . GENERAL INDUSTRY APPLICATION
* The Appendix F analysis indicates that relief valves may not be required in the majority of cases
* General agreement that valve leakage will occur to limit system pressures.
Lif ting of ADVs or solenoid valves                                .
Elastic stretch of bonnet studs Valve packing weepage
* The accident scenario of concern is lovi probability. Leak before break criteria will limit containment temperatures
* Most penetrations of concern are water filled and are not functionally used post accident
* The relief valves introduce long term testing, maintenance, and ALARA constraints
* The Appendix F approach is consistent with recent ASME Section lll Working Group recommendations
* The Appendix F option is the cost effective and the logical approach to resolve this industry issue I
Page 9 of 9 i
 
4
:                                                          i Risk Considerations o GL 96-06 Implementation I
l
!          Greg Krueger Generic Letter 96-06' Workshop
;        December 4,1997
                                          'W          *)
                                      , i i; 9 .. a,.y 4
 
i    Generic Letter 96-06 Risk l    Considerations    x __,_ .
l                                            _
;    e A probabilistic evaluation of comparative
)      risks was performed.
l
        + Probability of a containment integrity breach 4
due to an overpressurization event as described in GL 96-06 was evaluated.      .            j
,      + The likelihood of an inadvertent failure of a
:        relief valve requiring plant shutdown progressing to core damage was evaluated;            .l e .c i
 
Evaluating the Probability ofthe                        l .
Overpressurization Event m The likelihood of a containment integrity breach is composed of:
  + LOCA Frequency
  + Likelihood of system isolation
  + Probabi!ity of containment isolation valves remaining leak tight      .
  + Probability of pipe rupture occurring at specific locations
 
Determining Plant Operational 1                                Risk                      _
i l
;                                a Likelihood of core damage due to manual shutdown is composed of:                                                                                  '
t
                                    + Likelihood of relief valve opening i                                    inadvertently during power operation.                                                                          ;
i i
                                    + Conditional core damage frequency (CDF)                                                                        .
due to manual shutdown.
i
                                                                                                +                                                  .
j i '- l il (:4' i a [
                                                                                                                                                    -l
 
i                                                                                .
l Comparative Risk Conclusions
                . .- c - --~--- . ~ = ,=-
l m Risk associated with a required plant                                          '
shutdown due to inadvertent relief valve l  opening is increased.
i                                                                            .
m Probability of a containment integrity                                          '
!  breach due to system
!  overpressurization resulting from a                      -
i design basis LClCA is low.
i
                                                                    .a I  p] j, ],5% ''[
 
                                                  . O RiskEvaluation of Generic Letter 96-06 Oyster Creek Nuclear Generating Station l
GPU Nuclear Corporal: ion Ken Canavan - Safety & Risk Analysis
 
Purpose ofthe GL 96-06 RiskEvaluation l
Potential for early shutdown of Oyster Creek in the year 2000.
Some projects may have limited cost-benefit.
Therefore:                    DeferProiects With                  _
LimitedSafetyImpact UntilFinalDecision.
b
 
                                                  ~
Scope ofthe GL 96-06 RiskEvaluation Many projects evaluated for deferral.
Screening approach used to reduce number.
Detailed risk evaluation produced final request.
Deferral ofimplementation of(}L 96-06 requested until-18R refueling outage (year 2000) as part of integrated schedule change.
 
s 4
l Method ofthe GL 96-06 RiskEvaluation
~
i
~
Assumes that implementation is not complete.            ,
i Evaluates the effect on "Large Early Release i  Frequency" or LERF.
l  Several cases evaluated from the most bounding to more realistic.                '              '
Evaluation stopped, when or if, acceptable results reached.                                  -
                                                          .i
                                                          ~
i
 
O
-                                                            ~
Results ofthe GL 96-06 RiskEvaluation Case 1: Assumes ~all loss of coolant accidents (LOCAs) in the drywell result in piping failure which fails containment integrity.
    - Case 2: Assumes all Large LOCAs in the drywell result in piping failure which fails containment integrity.
    - Case 3: Assumes all Large LOCAs in the drywell result in a 50% chance of a failure of containment integrity.
l
 
\
Results o the GL 96-06 RiskEvaluation (continued)
Case 1: All LOCAs
        - Bounding case
        - LERF increase of 2.6x10-7 per year
        - Percent increase of 34%
        - Falls into the "Further Evaluation Required" category 1
(EPRI PSA Applications Guide) '
.I
;                                                                              -l
__    _    . . _ . -                                - _  -_____--_______]
 
                                                            ~
Results ofthe GL 96-06 RiskEvaluation (continued)
Case 2: All Large LOCAs
    - Less conservative case
    - Small LOCAs excluded since depressurization or containment spray will mitigate containment heatup.
    - LERF increase of 1.1x10-7 per year
    - Percent increase of 15%
    - Falls into the "Non-Risk Significant" category (EPRI PSA Applications Guide) l
                                      --c            m-
 
Results ofthe GL 96-06 RiskEvaluation (continued)
Case 3: All Large LOCAs Result in 50% Chance                                                        ,
l  of Containment Integrity Failure
  - Least conservative case
;                                                                                                      i
  - Small LOCAs excluded
!  - 50% chance of failure based on specific break location required and ultimate pipe failurei pressures
  - LERF increase of 5.5x10-8 per year
  - Percent increase of 7.5%
  - Falls into the "Non-Risk Significant" category'
                                                                                                    -I
__ j
 
O Conclusions ofthe GL 96-06 RiskEvaluation                                        \
Case 3 best represents the risk impact of the deferral of 96-06 implementation.
Conservative Large LOCA frequencies, estimation of pipe break locations and ultimate pipe failure pressures ensure margin in risk evaluation.                                                                ,
Risk impact of Generic Letter 96-06 deferral is
                              " Low".
 
o e B.! Generie Letter 96-06 Modifications The generic letter questions the operability of systems with regard to their cepability to withstand ambient heating followleg a loss of coolant accident (LOCA). Preliminary analysis indicates that two of the three issues contained in the generic letter do not apply to Oyster Creek (reference 6). The third issue has been determined to apply to Oyster Creek.
Project Description and Proposed Change The third issue is the overpressurization of containment penetrations due to ambient heating following isolation during a LOCA Without overpressure protecuon, the concern is that entrapped water between the inboard and outboard Isolation valves is heated, expands, and increases in pressure challenging the strength of the particular penetration. Five (5) penetrations require modification to relieve overpressure:
: 1.      Reactw Building Closed Cooling Water (RBCCW) Retum from the Drywell
: 2.      Shutdown Coeling Supply to the Reactor
: 3.      Isolation Condenser "A" Condensate Retum
: 4.      Isolation Condenser *B" Condensate Return
: 5.      Recirculation Sampling Operability determinations have been performed indicating that all systems considered susceptible to overpressure are operable for the interim duration until ( .er procedural changes and'or hardware modiG ations can be made (reference 6,7,8) GPU has committed to perfonn corrective actions which involve physical modifications to the plant be documented in the Integrated Schedule for Oyster Creek, pursuant to license condition 2.C (6) of the Full Term Operating License.
Risk Impact Evaluation ne analysis of the risk impact of the deferral of the 96-06 modifications until the 18R outage is performed using insights developed in the Level 1 and Level 2 OCPRAs. The Generic Letter 96-06 is primarily concemed with the integrity of the containment following a LOCA. That is, the overpressure of containment penetrations resulting in failure of the penetration and containment integrity. The figure of merit or risk measure used in the determination of the risk impact of the 96-06 modifications is, therefore, Large Early Release Frequency. Three cases are used in the estimate of the risk impact. The cases are ordered from most conservative to least conservative.
Senntivq Case 1 In Case I, the risk impact is estimated by assuming that all LOCAs which discharge to the drywell and result in core damage, also result in overp essurization and failure of a containment penetration. This includes the effr' of small LOCAs even though small LOCAs would not result in the severe environmental conditions that occur during large LOCAs. The contribution of all LOCAs to +be total core damage frequency is taken from the Level i OCPRA. Table B-1 provides the contribution of all LOCAs to the total core damage frequency.
Since it is assumed that piping overpressure results in the failure of a containment penetration, a luge containment bypass is created. This is a conservative assumption since a large containment bypass requires either of the following to occur: (1) a single large pipe rupture at the containment penetration or (2) two pipe breaks with one inside the containment and another outside.
The contribution of the LOCAs is normally an
* containment intact" plant damage endstate. To model the assumed containment integrity fail.tre, the normal plant damage endstate of " containment intact" is
 
e    e adjusted to large early release endstate. His leads to increase in the total large early release frequericy approximately equal to the core darnage frequency of the LOCA contributions. A "Large Early Release Frequency Worksheet" (LERF) is provided as Table B 2 and displays the estimation of the increase in LERF.
From Table B 1, the frequency of all LOCAs with discharge to the primar/ containment airspace is equal d
to 2.59x10 per year. In Table B 2, the Level 1 Key Plant Damage State PIFW, which is a " containment intact" endstate, is reduced by the above LOCA frequency of 2.$9x104 per year.
Key Plant Damage State PIFW (Base Case)- LOCA Frequency Contribution = New PIFW 1*fDS 4 4 1.16x104-2.59x10 = 8.98x10 The percent variance on Table B 2, is then 8.98x104 divided by the base case of 1.16x104 or 22%.
Also on Table B 2, the Level I Key Plant Damage State, MKCU, which is a large early release containment endstate, is increased by the above LOCA frequency of 2.59x104 per year.
Key Plant Damage State MKCU (base case) + LOCA Frequency Contribution = New MKCU KPDS d
1.72x104 + 2.59x1O = 4.31x104              ,
ne percent variance on Table B-2,is then 4.31x104 dhided by the br.se case of 1.72x103 or +151%.
The changes to these key plant damage states results in an increase of the large early release frequency 4                                      4 from the base case of 7.56x10 per year to 1.02x104 per year or 2.59x10 per year. This corresponds to an increase in the large early release frequency of 34.3%.
The Case I analysis of the risk impact of the deferral of the Generic Letter 96-06 modifications remains bounding due to the conservative assumptions regarding pipe rupture, pipe rupture locations as well as the assumption that small LOCAs result in the overpressurization of the susceptible containment penetrations.
SensitMry Case 2 Case 2, a less conservative sensitivity case, is evaluated to estimate a less conservative risk impact. This sensitivity case evaluates the risk impact assuming that the issue of piping overpressure is restricted to the large LOCAs into ths drywell airspace which result in core damage. That is, small LOCAs result in a lets severe environment due to the slower heatup of the drywell. The slower heatup allows for the initiation of containment spray and'or the automatic depressurization system. The effect of the cooling of the containment spray system and use of the automatic depressurization system to remove heat to the torus and results in less heat being discharged to the drywell. With less ambient heatup of the drywell and, therefore less ambient heatup of piping penetrations, it is less likely that piping failures due to overpressurization will occur. Using the *Large LOCA with discharge to the drywell airspace" row from Table B 1, the evaluation performed in case 1 above is repeated. The results are displayed on Table B-4.
4 In the sensitivity case, the increase in large early release frequency is 1.12x10 per year which corresponds to a 14.8% increase, This evaluation remains conservative due to assumptions with regard to assumed pipe breaks following exceeding code allewable stresses and the assumed break location (or multiple breaks) which fail containment integrity.
Sensitivny Case 3 In case 3 the risk impact is evaluated by assuming that piping overpressure is restricted to the large LOCAs into the dryw ell altspace which result in core damage and only 10% of piping overpressurizations result in
 
e                    e a pipe break which fails containment isolation. This is reasonable assuming that ultimate failure pressures of pipes are typically significantly higher than the design or code allowable pressares. For the total of five penetrations this is equal to 5 times 10%, or a 50% chance of containment integrity failure due to pipe overpressurization.
ne effect of small LOCAs is also not considered in this case. As stated in the , evaluation of case 2, small LOCAs result in a less severe environment due to the slower heatup of the drywell. Tne slower heatup allows for the initiation of containment spray and'or the automatic depressurization system. The cooling effect of the conuinment spray system and use of the automatic depressurization system to remove hew to the torus, results in less heat being discharged to the drywell. With less ambient heatup of the dr>well and piping penetrations it is less likely that piping failures due to overpressurization will occur. The frequency used is 50% of the frequency in case 2.
In this sensitivity case, the increase in large early release frequency is 5.5x10-' per year which corresponds to a 7.4% increase.
Results and Conclusions he results of three sensithity cases used to eYaluate the af eCl of the deferral of Generic Letter 96-06 Modifications is displayed on Table B.5, below. As the results indicate, the large early release frequency ranges from a percent increase of 7.4% to 34.3%.                                      .
Increases in the Large Early Release Frequency are categorized according to the criteria on Table 2 (found in the main report). The risk impact of the sensitivity cases range over risk categories of Low and Medium.
Based on judgement, case 3 is deemed to best represent the deferral of Generic Letter 96-06 Modifications and the risk impact is categorized as low.                                                                    ,
Table U.S- Summary of Generie Letter 96-06 Risk Evaluations Case Description                  Large Early Release Frequency          Risk Category increase          Percerit Value          Increase Case 1: All LOCA core damage frequency      contributions        (which      2.59x10 4          34.3 %          Medium discharge to drywell airspace) result in containment integrity failure.
Case 2: All Large LOCA core damage frequency contribunons (which discharge                      #
1.12x10            14.8%          Medium to the drywell airspace) result in containment integrity failure.
Case 3: 50% of Large LOCA core damage frequency contributions (which                5.5x 10''          7.4%              Low discharge to the drywell airspace) result in containment integrity failure.
N
 
i TABLE B 1 OC?RA INITIATING EVENT IMPORTANCE J                                                    MUUbL Name. UUPRA 13 Initiator Contributions to End State Group : ALL Total Frequency for the Group = 3.7982E-06 initiator                                                                                      #
t-requency      unaccountea i    Percent LU5F                                                                            1.24 h-05        1.006-09      32.73 %
TTRIP                                                                          4.64 E-07        3.64 E-09      12.23 %
RT                                                                              2.84 E-07        1.80E 09        7.48%
LOFW                                                                            2.60E-07          1.40E 09        6.85 %
CMSIV                                                                          2.57E-07        3.18E-09        6.76 %
LOTB                                                                            1.48E 07        9.61E 10        3.90 %
LOCV                                                                            1.48E-07        2.69E-09        3.99 %
LOIS                                                                            1.22E.07        7.26E 10        3.21 %
EPRL                                                                            1.19E-07        2.54E-09        3.13%
e v+;.p h e.c,;.m a. a pw.
m
                          -- __.                                -                :ws_.,i.l g
W. ween w rabcY-
                                                                                                "f69E'-O'9~ ~~~f65'I
  = uu/p. v-r.Wsne Ws.cxtewp.iug gus++0%                                .
_wgti  Wh 6.? ts itsiAV                                                                          9.04 E-68        C67E-09        2.38 %
PLOFW                                                                          7.83F-08          1.89E      2.06 %
LBIO                                                                            7.65E-08        2.86E 11        2.01 %
usei v.m eq ! ;.or.cm ,a. v,w. - - w                                                w    .a                              e Sf0                                                                            4.57E-08      1-gN.aw SE-11 e w p1.2 69.n .:
LOlA                                                                            3.15E-08          1.96E-09        0.83%
EPRH                                                                            2.93E-08          1.52E-09        0.77 %
LOCW                                                                            2.15E-08          1.35E-09        0.57 %
IADS                                                                            1.59E-08        9.01 E-11        0.42%
SAOTB                                                                          2.28E-09        4.89E 11        0.06 %
  $'Wts IXO h        bhM                                                      482M 10
                                                                                            $. ihi 3.36's 12 d
0.01 %
LAOIC                                                                          3.99E 11        3.36E-12        0.00 %
SAORB                                                                          2.66E-11          1.03E 11        0.00 %
i  SAOIC                                                                          8.52E-12          1.24E-11        0.00 %
IUIALS                                                                        3.5Uh-UU        Z.W5t-05        100.00%
n -  n- s
          ,    a.F; m                                                                          al "'h* $ higi
                                                                                                                      - "i ,%
n di Rid &d
 
l                                                                                                                                            1
?                                                                                                                                            l l
k I                                                      Table B 2: 96-06 LERF Estimation Spreadsheet Ref erence Case: Base Case (Risk Model: OCPRA 1 Case under study: 96 06 LERF (Case 1)
Level 1. Initiating Events                          Level 1 Key Plant DamaDe States
]                                    I E. Value      Reference    Vanance E        Input          input        Reference  Percent
!                              CMSIV        4.17 E-01    4.17 E-01      0%      B      KPDS          Value        Base Case  Variance EPRH          5.61 E-02    5.61 E-02      0%            PlFW        8.9 B E-0 7      1.16 E-06    22 %
EPRL          1.76 E-01    1.76 E-01      0%            NIFW        1.04506          1.04 E-06      0%
;                              LADS          1.33 E-03    1.33 E-03      0%            OlAU        5.75 E-07      5.75 E-07        0%
!                              IEMRV        3.31 E-02    3.31502        0%            OJAU          1.83 E-07      1.83 E-07        0%
l                              LAICS        6.21 E-05    8.21 E-05      0%            MKCU          4.31 E-07      1.72 E-07    151 %
!                              LAIMS        1.15 E-04    1.15 E-04      0%            MJAU          5.88508        5.88 E-08        0%
,                              LAOIC        6.96 E-08    6.96 E-08      0%            NJHW          1.56 E-08      1.56 E-08        0%
LAOMS        6.44 E-08    6.44 E-0 8    0%        Total CDF        3.80 E-06      3.80 E-06        0%
;                              LBI          5.67 E-04    5.67 E-04      0%                                                                4 l                              LDIO          8.37506      8.37506        0%                              LERF Estimation l                              LOCV          2.24 E-01    2.24 E-01      0%                        Percent of CDF Analyzed =      '4.3 6 %
l                              LOCW          2.71 E-02    2.71 E-02      0%                        Total analyzed frequency =  3.20 E-06 LOFC          1.71501      1.71 E-01      0%        Category 1 A . Large Entry LOFW          1.51501      1.51 E-01      0%            MKCU            100%          4.31 E-07  4.3' 507 LOlA          4.33 E-02    4.33 E-02      0%            NIFW          30.85%        1.04 E-06  3.22 E-07  ,
LOIS          7.51 E03      7.51 E-03      0%            OIAU          0.95%          5.75E-07  5.47509 LOSP          3.26 E-02    3.26502        0%                                            Total  7.58 E-07  :
LOTB          1.03502      1.03 E-02      0%                      Percent of Total Analyzed =  23.65%      l PLOFW        1.78 E-01    1.78 E-01      0%        Category 1B . Containment Bypass RT            7.21 E-01    7.21501        0%            OJAU            100%          1.83 E-07  1.83 E-07 sal          9.27 E-03    9.27 E-03      0%-          MJAU            100%          5.88508    5.88 E-08  -
SAOIC        1.59506      1.59 E-06      0%            NJHW            100%          1.56 E-08  1.56508    i SAORB        7.70 E-07    7.70 E-07      0%                                            Total  2.57 E-07 SAOTB        3.64 E-04    3.S4 E-04      0%                      Percent of Total Analyzed =    8.03 %
SBl          7.81 E-03    7.81 E-03      0%        Total LERF (sum of above) =                1.02 E-06 SBO          2.86506      2.86 E-06      0%                                  Reference LERF =  7.56 E-07 TTRIP        8.97 E-01    8.97 E-01      0%                        Percent Change in LERF =    34.26% E Percent of Total Analyzed =    31.69 %
EPRI PSA Applications Guide Risk significant cutoffs:                Risk Significant Cutoff          Delta for this Case                  ,
CDF              51.40 %                          0.00%                        ,
LERF              36.37 %                        34.26 %
l                              Comments:
i
  --,-.,--..-.ww,.-.vww.w-.w.~                                                                                                      -
 
!                                                  Table B 3: 96-06 LERF Estimation Spreadsheet I
f                                                Reference Case: Rose Case Risk Model: OCPRA 13) j                                              Case under study: 90-06 LERF (Case 2)
!                                        uvel 1. Initiating Events                                                    Level 1 Key Plant Damage States 1                  E          l.E          Value      Reference                    Vanence E                  Input        input              Reference        Percent l                          CMSIV          4.17 E-01    4.17601                              0%          E    KPDS          Value            Base Case        Variance j                          EPRH          5.61 E-02    5.61 E-02                            0%                PlFW        1.04 E-06              1.16 E-06        10%
l                          EPRL          1.76501      1.76 E-01                            0%                NIFW        1.04506              1.04 E-06          0%      l l                          IADS          1.33 E-03    1.33 E-03                            0%                OIAU        5.75 E-07            5.75 E-07          0%
!                          IEMRV          3.31 E-02    3.31502                              0%                OJAU        1.83507              1.83507            0%
!                          LAICS          8.21 E05      8.21 E-05                            0%                MKCU        2.84 E-07            1.72 E-07        65%      ,
2 LAIMS          1.15 E-04    1.15 E-04                            0%                MJAU        5,68 E-08            5.88 E-08          0%
LAOIC          6.96 E-08    6.96 E-08                            0%                NJHW        1.56 E-08            1.56 E-08          0%      ,
,                          LAOMS          6.44 E08      6.44 E-08                            0%            Total CDF      3.80 E-06            3.80 E-06          0%
LBI            5.67504      5.67 E-04                            0%
LBIO          8.37506      8.37 E-06                            0%                                LERF Estimation LOCV          2,24 E-01    2.24 E-01                            0%                          Percent of CDF Analyzed =                84.35 %
LOCW          2 71 E-02    2.71 E-02                            0%                          Total analyzed frequency =            3.20506 LOFC          1.71501      1.', i E-01                          0%            Category 1A Large Esify LOFW          1.51 E01      1.51501                              0%              MKCU          100 %              2.84 E-07      2.84 E-07 LOlA          4.33502      4.33502                              0%              NIFW          30.85 %              1.04506        3.22 E-07 l                          LOIS          7.51 E-03    7.51 E-03                            0%              08 " 1        0.95%                5.75 E-07      5.4 7 E-09
!                          LOSP          3.26 E-02    3.26 E-02                            0%                                                        Total    6.11 E-07 1
.                          LOTB          1.03 E-02    1.03502                              0%                        Percent of Total Analyzed =              19.07 %
PLOFW          1.78 E-01    1.78501                              0%            Category 1B . Containment Bypass RT            7.21 E-01    7.21501                              0%              OJAU          100%                1.83 E-07      1.83 E-07 SAI            9.27 E-03    9.27 E-03                            0%              MJAU          100 %              5.88508        5.88508 SAOIC          1.59 E-06    1.59 E-06                            0%              NJHW          100 %              1.56 E-08      1.5C E-08 SAORB          7.70E07      7.70 E-07                            0%                                                        Total    2.57507 SAOTB          3.64 E-04    3.64504                              0%                        Percent of Total Analyzed =              8.03%
SBl            7.81 E-03    7.81 E-03                            0%            Total LERF (sum of above) =                          8.68 E-07 SBO            2.86606      2.86 E-06                            0%                                    Reference LERF =            7.56 E-07 TTRIP          8.97 E-Oi    8.97 E-01                            0%                          Percent Change in LERF =          l 14.82% E Percent of Total Analyzed =              27.10%
W Risk, significant cutoffs:                                                                          Delta for this Case CDF                                        51.40 %                      0.00%
LERF                                        36.37 %                      14.82 %
l                          Comments:
 
I I
i Table B 4: 96-06 LERF Estimation Spreadsheet                                                  I Ref erence Case: Base Case              Model: OCPRA.13            -
l                                                          Case under study: 96 06 LERF (Case 3) i Level 1. Initiating Events                                          1 evel 1. Key Plant Damage States
: 1. E.        Value                    Reference    Variance E          Input          input        Reference      Percent I.                                CM SIV        4.17 E-01                  4.17 E-01      0%      E        KPDS            Value        Base Case      Variance EPRH          6.61 E-02                    5.61 E-02      0%                PIFW        1.10 E-06        1.16 E 06        5%
EPRL            1.76 E-01                  1.76 E 01      0%              NIFW          1.04 E-06        1.04 E-06        0%
LADS          1.33 E-03                    1.3 3 E-03    0%              OIAU          5.75 E-07        5.75 E-07        0%
IEMRV          3. 31 E-02                  3.31 E-02      0%              OJAU          1.83 E-07        1.83 E-07        0%
1.AICS        8.21 E 95                  8.21 E-05      0%              MKCU          2.28 E-07 l                                                                                                                                            1.72 E-07      33%
LAIMS          1.15 E-04                    1.15 E-04      0%              MJAU          5.8 8 E-08      5.8 8 E-08      0%
LAOIC          696 E-08                    6.96 E-08      0%              NJHW          1.56 E-0 8      1.56 E-08        0%
I                                LAOM S        6.44 E-08                  6.44 E-08      0%            tal CDF          3.80 E-06        3.80 E-06        0%
LBI            5.67 E-04                  5.67 E-04      0%
LBIO          8.37 E-06                  8.* 7 E-06      0%                                LERF Fatimation t.OCV          2.24 E-01                  2.24 E-01      0%                          Percent of CDF Analyzed =          84.35 %
l                                LOCW          2.71 E-02                  2.71 E-02      0%                          Total analyzed freauency =      3.2 0 E-06 i j                                LOFC          1.71 E-01                  1.71 E-01      0%          Category 1 A . Large Earfy LOFW i 51 E-01                  1.51 E-01      0%              MKCU            100%          2.2 8 E-07    2.28 E-07  '
!                                LotA          4.3 3 F-02                  4.33 E-02      0%              NIFW          30.85%          1.04 E-06    3.22 E-07
)                                LOIS          7.51 E-03                  7.51E 03        0%              OlAU          0.95%          5.75 E-07    5.47 E-09 l                                LOSP          3.26 E-02                  3.26 E-02      0%                                                Total      5.55 E-07 l                                LOTB          1.0 3 E-02                  1.03 E-02      0%                        Percent of Total Analyzed =        17.32 %
j                                PLOFW          1. 7 8 E-01                1.78 E-01      0%          Categor,v 18. Containment Bypass
)                                RT            7.21 E-01                  7.21 E-01      0%              OJAU            100%          1.83 E-07    1.83 E-07
)                                sal            9.2 7 E-03                  9.27 E-03      0%              MJAU            100%          5.88 E-08    5.88 E-08
)                                SAOIC          1.59 E-06                  1.59 E-06      0%            *!J HW            100%          1.56 E-0 8  1.56 E-08 i
SAORB          7.7 0 E-07                  7.70 E-07      0%                                                Total      2.57 E-07 S/sOTB        3.64 E-04                  3.64 E-04      0%                        Percent of Total Analyzed =        8.03%
SBl            7.81 E-03                  7.81 E-03      0%                                                          8.12 E-07 Total LERF (sum of above) =
i                                SBO          2.86 E-06                    2.86 E-06      0%                                    Reforence LERF =      7.56 E-07 TTRIP          8.97 E-01                  8.97 E-01      0%
Percent Change in LERF =      l 7.41 % E Percent of Total Analyzed =
25.35%
1                                                                                                                                A EPRI PSA Applications Guide l                                Risk significant cutoff s:                                R:sk Significant Cutoe              Deha for this Case l                                                                              CDF                51.40 %                            0.00%
!                                                                              LERF              36.37 %                            7.41 %
i i                                Comments:
I i
l i
a
 
e o GL 96-06 REM RENCE AIATERIALS INDUSTRY                                                                .
: 1. NEI Letter to NEI Administrative Points of Contact, Applicability of Generic Letter (GL) 91 18, Revision 1, Information to Licensees 1 Regarding NRC Inspection Manual on Resolution of Degraded and
* Nonconforming Conditions to Generic Letter (GL) 9G 06, Assurance of Equipment Operability and Containment Integrity During Design Basis Accident (DBA) Conditions, October 16,1997.
: 2. NEI Letter to NEI Nuclear Strategic Issues Advisory Committee, Generic Letter (GL) 96 06 Assurance of Equipment Operability and Containment Integrity During Design Basis Accident (DBA)
Conditions, November 7,1996,                      ,
NRC
: 3. NRC Generic Letter 96 06, Supplement 1, Assurance of Equipmec Operability and Containment Integrity During Design Basis Accident Conditions, November 13,1997.
i                      4. NRC Generic Letter No. 91 18, Revision 1, Information to Licensees l                          Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions,0ctober 8,1997.
: 5. NRC Letter to NEI, a summary of April 30,1997 meeting with NEI regarding Generic Letter (GL) 96 06, Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions, September 5,1997.
: 6. NRU Memorandum DRS Division Directors, Staff Guidance Associated wit!. Plant Startup Relative to Issues Addressed by Generic Letter (GL) 96 06, January 24,1997.
: 7. NRC Memorandum to David B. Matthews, A summary of December 19,1996 meeting with NEI regarding industry responses to Generic Letter (GL) 96 06, December 26,1996.
: 8. NRC Summary of the October 29,1996 meeting with NEI, November 22,1996
 
a ,
: 9. NRC Genenc Letter 96 06, Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions, September 30,1996,
: 10. NUREG 0933, A Prioritization of Generic Safety Issues, pages 910, June 30,1994.
: 11. NUREG 0933, Issue 150: Overpressuri ation of Containment Penetrations, Revision 1, June 30,1995.                          '
i
 
i        0 NEI NU(llAt (NitGY INSil1Uit
                                                                                                  -    DeM J. Modoen Mo"oeuEama, BY BROADCAST FAX October 16,1997                                                                                                '
TO:                  NEI Administrative Points of Contact
 
==SUBJECT:==
Applicability of Generic Letter (GL) 9118, Revision 1,Information to Licensees Regarding NRC Inspection Manual on Resolution of Degraded and Nonconforming Conditions to Generic Letter 96 06, Assurance of Equipment Operability and Containment Integrity During Design Basis Accident (DBA) Conditions                            .
Revision 1 to GL 9118 was issued on October 8,1997 and may pmvide a basis for revising licensee GL 96 06 corrective action implementation plans and schedules.
Please orovide this information to the individual in your organization ragonsible for imnlementine GL 96-06 commitments, GL 96 06 addresses two topics: (1) containment air coolar systems experiencing a water hammer or two phase flow, and (2) isolated water filled piping in containment experiencing thermally induced overpressurization. In response to GL 96 06, licensees generally identified one of the following approaches to address the potential for overpressurization:
* Implement plant hardware modifications to provide pressure relief. This typically requires the addition of pressure relief valves or expansion volumes to the piping system.
* Demonstrate compliance using the ASME Code Section III, Appendix F, Rules for Evaluating Service Loadings with Level D Service Limits. Most licensees would need to submit a license amendment to the NRC to add the appendix to the plant licensing basis.
Last spring, iflicensees were to use the approach outlined in Appendix F and apply for a license amendment, the NRC staff stated that an unreviewed safety question (USQ) per 10 CFR 50.59 was likely to arise. The NRC staff concluded, using the original guidance in GL 91 18. that the existence of a USQ would prohibit restart from an outage prior to the NRC approving the license amendment. This conclusion discouraged most licensees from choosing the Appendix F method to resolve an overpressunzation concern.
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    .                                                                                      . r NEI Administr:tiva Points of Contact October 16,1997
      - Page 2 Revision 1 of GL 91 18 revised the restart limitation in Section 4.8. It now states:
              ...the need to obtain NRC approval for a change (e.g., because it involves a USQ) does not aHect the licensee's authority to operate the plant. The licensee may make mode changes, restart from outages, etc., provided that necessary equipment is operable and the degraded condition is not in conflict with the TS or the license.
      - Schedule relief for GL 96-06 modi 5 cations may also be possible based on Section 4.3    i of the revised document. It states:
The licensee must establish a time frame for completion of corrective action.
The timeliness of this corrective action should be commensurate with the safety signi6cance of the issue.
Earlier today, NE1 met with NRC staff to discuss generic industry concerns relative
      --to implementing plant modifications as a result of GL 96-06 commitments. The -
_NRC staff expressed suyrise when we reported broad licensee concerns that many            ,
of the plant modifications underway or planned may contribute little to plant safety, yet be relatively costly to install. The NRC staff reported havmg received very few expressions of concern from individuallicensees. The NRC staff emphasized to us their willingness to consider requests for schedule relief as well as a change in the committed corrective action consistent with the guidance' contained in the recent revision to GL 9118.
You are encouraned to evaluate GL'91-18. Revision 1. to determine if.it
      - nrevides a basia.for revisinn any GL 08 06 corrective action commitments and schedules.
If you have questions about this in'ormation, please contact Kurt Cozens at (202) 739 8085 or koc@nei.org.--
        ~ Sincerely, b-    -                      _
David J. Modeen KD: nab -
c: NEI GL 96-06, Points of Contact NEI GL 96-06, Issue Task Force
 
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                                                                                      %wC68 As ot%gaanos BY_ BROADCAST FAX November 7,1996                                                                            -
TO:                NEI Nuclear Strategic Issues Advisory Committee (NSIAC)
 
==SUBJECT:==
Generic Letter 96-06 " Assurance of Equipment Operability and Containment Integrity Durmg Design Basis Accident (DBA)
Conditions" On September 30,1996, the NRC issued the subject generic. letter in order to:
(1) notify licensees about safety-significant issues that could affect contamment integrity and equipment operability under DBA conditions, (2) request certain information be submitted from licensees relative to those issues, and (3) request thailicensees implement actions as appropriate to address those issues. The purpose of this letter is to apprise you of NEI actions and plans in connection with this issue.
NEI hosted a meeting oflicensee representatives and NRC staff on October 29, 1996, to discuss questions concerning the generic letter. The NRC staffin attendance were those directly responsible for reviewing licensee responses to the genericletter. NRC staff answered questir os NEl had provided in advance of the meeting, as well as those raised during it. They plan to issue a meeting summary including answers to questions posed during the meeting, within three weeks.
The following key points were made during the discussion:
      . NRC staff views timely review and response to be very important. The emergency nature of the generic letter (e.g., being issued without the opportunity for public comment) was provided as rationale for NRC staff expecting that plants currently in a shutdown condition would atleast determme if their systems are susceptible to two-phase flow, water hammer, or overpressurization conditions, and if so, complete an operability evaluation prior to plent restart regardless of the response date requested in the generic letter of January 28, 1997.
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NSIAC November 7,1996 Page 2
      . NRC staff does not plan to review and approve all operability evaluations, but -
would expect such evaluations to be on file at the utility for review.
      . NRC staff has yet to determine a review plan or process.
Some licensee representatives believe the generic 1etter statements on pages 6
          - and 8 relative to U.S.A. Standarti B31.1 or ASME piping design code requirements for overpressure protection are incorrect. (Note: We are aware of discussions between a code committee member and NRC staff that should result in clar3 cation of the NRC staff position. We will promptly inform your staff of-any change.)
We appreciate the NRC's reasons for requesting licensee review and submittal of information relative to the issues reported in the generic letter. Severallicensee -
event reports indicated that cystem performans , under DBA conditionn may not be as analyzed in their FSAR. Yet, based on the technical insights discussed during the remainder of the industry meeting (noted below), there is strong evidence to suggest that despite the two-phase fluid conditions, containment heat removal and integrity would still be achieved.
Licensee representatives from five plants with event reports described in the genericletter presented their acsessment e.nd corrective action activities. The following technical insights are importan..
      . - Many plants have already determined that two-phase rather tnan single-phase
            -fluid conditions may occur during the hypothetical DBA. But most plants have -
not completed the engineering assessment of the significance of a two-phase fluid condition relative to system operability under DBA conditions.
Limited testing of scaled plant-specific system configurations by two licensees demonstrated only small transient loads are induced by the two-phase fluid condition and subsequent y, dem response. These results indicate that for the two plants analyzed, the em nment and containment air cooling system would meet their design requirements.
_ Plant-specific system design and layout vary considerably and make generic, bounding assessments difficult.
* Potential design modifications to preclude two phase conditions in contamment air coolers, especially for an open system design, can be complex and expensive.
O
 
NSIAC November 7,1996 Page 3 Based on these discussions, it appears further coordinated industry efforts could result in more cost-effective utility responses to the generic letter request. Thus, NEIis taking the following steps:
: 1. Establish an NEIissue task force to respond to the generic issues raised in the generic letter. We expect the task force to consist of representatives from licensees, EPRI,IMPO, applicable ash'E code committees, and an              .
architect / engineer.
: 2. Request that EPRI sponsor analytical and experimental work to evaluate the impact of the two-phase fluid condition in varying containment air cooler system Configurations.
: 3. Meet with NRC staff to ensure that relevant techuicalinformation resulting from generic industry work is well understood. We would use such meetings to '
explain any differing views industry may have regarding the validity of statements made in the generic letter and subsequent NRC staff written responses to the industry questions posed at the industry meeting. Obviously, some information would be available prior to the 120-day response date whereas other information may be developed later.
In summary, we believe that development and sharing of additional technical and regulatory information will assist licensees in providing responses to the generic letter that are more systematic, complete, and cost-effective. Your personal attention to this matter is important. We respectfully request that you:
Provide NEI (attention of Dave Modeen) a copy of your 30- and 120-day (due January 28,1997) response letters.
          .                                                                                      i Complete the attached survey form for each of your plants by November 20,
'            1996. The survey will provide critical parameters for grouping of plants with similar containment air cooling system configurations.
Support the EPRI effort to develop a more robust technical bases for analyzing and explaining the impact of two-phase fluid conditions and water hammer.
EPRI has a draft plan for the collaborative affort which can include non-members of EPRI. Interested parties should contact the EPRI project manager, Avtar Singh at 415-855-2384 or by e-mail at avsingh@msm.epri.com.
Any questions should be directed to Dave Modeen at N2) 739-8084 or by e-mail at djm@nci.org.
  )
1                        -
 
NSIAC November 7,1996
    - Page 4 Sincerely, 44    .
Ralph E. Beedle DJM/amj                                                                      -
Enclosure c:    NEI Administrative Points of Contact 9
 
    .  . _ _ __~                _        _ . _          _ __                                        ._      _    _ _ _ _ _ _ _ _ _ _ _ . _ _ .
t              4-Generic Letter 96 06                                                      ,
Containment Fan Cooler Evaluatirn Plant Classification Survey Utility:                                                                    Plant:                                    _
L l ~.JIR E E is the plant in an outage between now and January 28,19977                                                                -      l          l        l Are the containment fan coolers credited in the LOCAMSLB analyses (peak temperature and pressure)?                                                                                                l          l        l Ifyes:
Is the service water for the containment fan coolers in an open system? -                                            !          l        I Ifyes:
Are the fan coolers located on an elevation 2 32 feet above the lake / river water level?                                                                                l          l        l Ifno:
Is the surge tank at atmospheric pressure?                                                              l          l        l Ifyes:                                                                                            -
Are the fan coolers located on an elevation 2 32 feet above the surge tank water level?                                                          l          l        l Ifno:
i                                    Does the service water supply to the containment fan coolers isolate on SI signal?                                                                                                            L          I        I Ifyes:
Are relief devices provided for thermal expansion of trapped fluids:                                                                                                l          l        l Are there other heat loads (possibly isolated under accident conditions) the rely on the same service water system?                                                                                      l          l        .l Have you concluded the containment fan coolers are susceptible to the two-phase fluid condition?                                                                                                        l          l        l The following data will help to further classify fan cooler evaluation issues:
Maximum LOCA containment temperature                                                                                                              aF Fan cooler data:
Manufacturer and model Number of fan coolers Number of coils per tan cooler unit Number of tube passes Design fouling factor for fan cooler
.                                    Accident service water flow rate                                                                                                      com Accident service water inlet and discharE e temperature                                                                        /      *F What do you see as the next step to coordinate the industry response to this issue:
 
  .,    4 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555 0001 November 13,1997        _
NRC GENERIC LETTER 96-06, SUPPLEMENT 1: ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS Addreemans Alt holders of operating boonses for nuclear power reactors except those who have permanendy onesed operations and have certified that fuel has been permanently removed from the reactor vessel.
PJERGRt
          'The U.S. Nudear Regulatory Commission (NRC) is issuing this supplement to Generic Letter (GL) 96 06, '' Assurance of Equipment Operabitty and Centainment integrity During Design-        ,
Basis Accident Conditions," to inform addressees about ongoing efforts and new developments associated with GL 9606 and to provide additional guidance for completing corrective actions. Addressees may find this information useful in planning and scheduling future actions associated with GL 96 06. This gener'c letter suppiament contains no new NRC requirements. Furthermore, ne speel6c action or wntten response is required.
Backaround GL 96-06 war, issued on September 30,1996, to address the followinn issues of concom:
: 1. Cooling water systems serving the containment air coolers may be exposed to the hydrodynamic effects of waterhammer during either a loss-of-coolant accident (LOCA)
}                or a main steam line break (MSLB). These coolmg water systems were not designed to withstand the hydiMyr,eirik effects of waterhammer and actions may be needed to sabsfy system design and operability requirements.
: 2.      Cooling water systems serving the containment air coolers may expenonce two-phase flow conditions during postulated LOCA and MSLB scenarios. The heat fornovel assumptions for design-basis accsdent scenarios are based on single-phase flow conditions and actions may be needed to sabsfy cystem design and operability requirements.
: 3. Thermally induced overpressurization of isolated water-filled piping sections in containment could jeopardize the ability of accident-mitigating systems to perform their safety functions and'could lead to a breach of contamment integrity through bypass leakage. Actions may be needed to sabsfy system operability requirements.
      < J 465vv.i
 
l-l t
GL 96-06, Supplement 1 November 13,1997 Page 2 of 5 l                                            GL Of46 states-
                                                    "If systems are found to be suscepbble to the conditions 30W in this generic letter, addressees are expected to essess the operability of affected systems and take corrective action as appropriate in accordance with the requirements stated in 10 CFR Port 50 Appendix B and as requireef by the                    ,
f plant Technical Specifications."
GL 96 06 refers to GL 91-16, "Irdormebon to Licensees Regardng Two NRC inspechen Manual Sections on Resolution of Degraded and Nonconforming Conditions and on l                                            Operability," for guidance on the r==akslan of issues identlSed in GL 96-06. (Note:
GL 91-18, Revision 1 we 8*eued on October 8,1997, to inform licensees of the issuance of i-                                          a revised section of th6 WC Inspection Manual, Part 9000, ' Technical Guidance," or; the
;                                            resolution of degraded and nons,Jwi..;,@ condthons. The GL 96 06 reference to GL 91-18 5
remains valid since it refers or,1y to the " Technical Guidance" on operability, which was not changed by the issuance of GL 91-18, Revision 1.) Criterion XVI, " Corrective A7tions," of i
Appendix B to 10 CFR Part 50 states, in part, " Measures shall be established to assure                    ,
l                                          that...nonconformances are promptly identified and conected." _in this regard, GL 91-18 states that the timeliness of conective actions should be commensurate with the safety i                                          sientAcance of the issue, and that the conective action requirements of Appendix B may be
;                                          satisfied by making changes in the design of the plant in lieu of restoring the affected g
equipment to its original design. In one example, GL 91-18 spedfically discusses the use of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Sechon lil, Appendix F, criteria for interim operability determinsbons for degraded and nonce ,i-Tr.;rs piping and pipe supports. It states that the use of Appendix F critoria is valid
!                                          until the next refueling outage when the supports are to be restored to the final safety analysis report critoria.
i i                                          Addressees have responded to the generic letter and have estabhshed schedules for resolving the GL 96 06 issues. The NRC staff is cunently reviewing the information that has been submrtted.
I                                          Discussion i
                                        . Implementing conective actions to resolve the GL 9606 issues can have a signl6 cant impact
;                                          on outage schedules and resources, and some addressees have indicated that it would be prudent to take more time to better understand the ep.w#,c concems that have been idenbfled in order to optimize whatever modificebons are needed and to assure that they do not ultimately result in a detriment to safety. Current 'asues and ongoing efforts that could
;                                          inficace en addressee's decision in planning conective actions include: (1) risk im,pi A,s                    :
of installing relief valves to dea! with the thermal overpressurization issue; (2) feasibility of
;                                          using the acceptance critoria contained in Appendix F to Section 111 of the American Society j                                          of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for the permanent 1
resolution of the GL 96-06 issues; (3) ongoing tests by the Electric Power Research Institute to support a generic resolubon of the overpressunzation of piping issue; and (4) queshons                    '
 
GL 96-06, Supplement 1 November 13,1997 Page 3-of 5 regarding the staff's closure of Generic Safety issue 150, "Overpressunzabon of Containment Penetrabons." Risk insights and industry inibatives that are being considered or that may be proposed could also influence the course of action that addressees take to resoh's the GL 9646 issues.                                                      ~
Addressess are responsible for assessing equipment operability, determining acbons, and r-i"l#5 schedules that are appropriate for resoMng the specific conditions that have been idenkfled. in determining the appropriate actions and schedules for resoMng GL 9606 lasues, addressess should consider, for example, the continued validity of exishng operability detemiinations, compensatory edions required to maintain operability, the safety significance associated with the specific nonconformances or degraded conditions that have been identifled, risk insights, and the time required to complets any generic initiatives and/or plant-specific actions (e.g., engineering evolustions, design change packages, material procurement, and equipment modification and installation). Also, analytical solutions employmg the permanent use of the acceptance criteria contai,wd in the ASME Code, Section ill, Appendix F (or other acceptance criteria) may present viable altamabves to plant modifications and can t'e used where appropnate, justified, and evalua*-d in accordance with NRC requirements such as 10 CFR 50.59, as applicable. Addressees may find the revised guidance contained in GL 91-18, Revision 1, dated October 8,1997, helpful in determining i
appropnate actions and schedules. ANhough adjustments in schedules may be warranted on i
the basis of these (and other) considerations, specific adten met have been ifhed and are clearly needed should not be delayed without suitable justifhation.
It is the staff's cunent position that addressees can use the ASME Code, Section 111, Appendix F criteria for interim operability determinations for degraded and nongrifwks piping and pipe supports until permanent actions have been identifled and approved by the NRC (as applicable) for resolving the GL 96-06 issues. This guidance suppiaments the guidance provided by GL 91-18 for resolution of the GL 96-06 issues, in order to further fac4 tate resolution of the GL 96-06 issues, the NRC will parbespete in a public workshop scheduled for December 4,1997. The worinhop proceedings will be -
summarized by the NRC staff and made putdicly available. The need foi additional NRC guidance and generic communication will be consulerod upon cranplebon of the workshop.
Reauested Infomutton Addressees who choose to revise their commitments for resolving tlw GL 96-06 issues should submit a revised response to the genenc letter. Revised repwes should include appropnate diam *=ian of the considerabons discussed abcve, the cummt resolution status and achons remaining to be completed, and plans being considered for final resolution of the GL 96-06 issues.
_9
 
                                                                                                =. ,
GL 36 06, Supplement 1 November 13,1997 Page 4 of 5 Federal Remater Notification Because this GL supplement is informational, requires no specific acbon or response, ano is the result of ongoing efforts between NRC staff and addressees to resolve GL 9606 issues, there is no need for edchbonal opportunities for comment. Accordingly, a notim of opportunity for public comment was not published in the Federal Regisfer. However, comments on the content of this supplement to GL 96-06 may be sent to the U.S. Nuclear Reguistory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 0001, Panenwork Radudion Act Statement For thans addressess who find it necessary to revise the'r commitments for n "As the GL 96-06 isses, this generic later supplement contems mformabon coiledions that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3601 et seq.). These information colisations were approved by the Office of Management and Budget, approval number 3150-0011, which expires on September 30,2000.
The public repoiting burden for addressess who find it necessary to revies their response to GL 3606 is estimated to average 40 hours per rosponse, including the bme for reviewing instructions, searching avi=*ing data sources, gathering and maintaining the dets needed, and completmg and reviewmg the collection of information. The U.S. Nucieer Regulatory Commission is seeking public comment on the potentialimpact of the collection of information contained in the poneric letter and on the following issues (1)    la the picpc;;d collection of information necessary for the proper performance of the furxtions of the NRC, induding whether the informahon will have practical utlBty?
(2)    is the eshmate of burden accurate?
(3)    is there a way to enhance the quality, utility, and clarity of the informebon to be collected?
(4)    How con the burden of the collection ofinformation be minimtzed, including the use of automated collechon techniques?
Send comments on any aspect of this collechon of information, including suggestions for e
reducog this burden, to the information and Records Management Branch, T 6F33, U.S.
Nut.%ar Regulatory Commission, Washington, D.C. 20555-0001, and to the Desk Officer, Offica of information and Regulatory Affairs, NEOB-10202 (3150 0011), Office of Manaaement and Budget, Washington, D.C. 20503.
 
o
* GL 96-06, Cupplement 1 November 13.1997 Page 5 of 5 If you have any questions about this matter, please contact the lead project manager or one of the te&nical contacts listed below, or the appropriate Of50s of Nuclear Reactor Regulation (NRR) project manager for a speer6c nuclear power plant.
'                                                          g Jack W. Roe, Ading Director Division of Reactor Program Management Of5ce of Nucear Reactor Regulation Technical contacts: James Tatum, NRR 301-415-2405 E-mail: jotignrc. gov John Fair, NRR 301 415-2759                                    .
E-mail: jrf@nrc. gov Lead Project Mana0er: Beth Wetzel, NRR 301 415 1355 E mail: bew9nre. gov Attachment List of Recently issued NRC Generic Letters 1
 
                                                                                    +      o Ar. .?. ment GL 96-06, Suppiament 1 November 13,1997 Page 1 of 1 UST OF RECENTLY ISSUED GENERIC LETTERS Generic                                Date of Letter    Itghet                        lastrace        leW To 91-18,  INFORMATION TO LICENSEES        10/08/97        ALL HOLDERS OF OLs REV,1 REGARDING NRC INSPECTION                            FOR NUCLEAR POWER MANUAL SEC110N ON RESOLUTION.                    AND NPRs, INCLUDING OF DEGRADED AND NONCONFORM-                      THOSE POWEP AEACTOR ING CONDITIONS                                    LICENSEES WO HAVE PERMANENTLY CEASEC OPERATIONS, AND ALL HOLDERS OF NPR LICENSES WHOSE LICENSE NO LONGER AUTHORIZES OPERATION 97-04  ASSURANCE OF SUFFICIENT          10/07/97        ALL HOLDERS OF OLs NET FOSITIVE SUCTION                              FOR NUCLEAR POWER HEAD FOR EMERGENCY                                PLANTS, EXCEPT THOSE CORE COOLING AND                                  WHO HAVE PERMANENTLY CONTAINMENT HEAT                                  CEASED OPERATIONS AND REMOVAL PUMPS                                      HAVE CERTIFIED THAT FUEL HAS BEEN PERMAN-ENTLY REMOVED FROM THE REACTOR VESSEL 97-03  ANNUAL FINANCIAL SURETY          07/09/97        URANIUM RECOVERY LICENSEES UPDATE REQUIREMENTS                              AND STATE OFFICIALS FOR URANIUM RECOVERY                                                      ,
LICENSEES 97-02    REVISED CONTENTS OF              05/15/97        ALL HOLDERS OF OLs THE MONTHLY OPERATING                            FOR NPRs, EXCEPT THOSE REPORT                                            WHO HAVE PERMANENTLY CEASED OPERATIONS AND HAVE CERTIFIED THAT FUEL HAS BEEN PER-MANENTLY REMOVED FROM THE REACTOR VESSEL OL = OPERATING LICENSE CP = CONSTRUCTION PERMIT NPR = NUCLEAR POWER REACTORS j
TOTAt_ P.07 h                          _
 
9 0 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001 October 8,1997 NRC GENERIC LETTER NO. 91 18 9EVISION 1: INFORMATION TO LICENSEES REGARDING NRC INSPECTION MANUAL SECTION ON RESOLUTION OF DEGRADED AND                                                  .
NONCONFORMING CONDITIONS Addressees All holders of operating licenses for nuclear power and non power reactors, including those power reactor licensees who have permanently ceased operations, and all holders of non-power reactor licenses whose license no longer authorizes operation.
Purpose The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to inform licensees of the issuance of a revised section of Part 9900, " Technical Guidance,' of the NRC Inspection Manual. The revrsed section is entitled " Resolution of Degraded and Nonconforming Conditions.' The revisions to this sectron of Part 9900 more explicitly discuss the role of the 10 CFR 50.59 cvaluation process in the resolution of degraded and nonconforming conditions. The Part 3900 guidance on operability forwarded by Generic Letter (GL) 91-18 has not been revised. This letter is provided for information only, no specific action or written response is required.
Backaround The previous version of NRC inspection Manual, Part 9900, " Technical Guidance,' on the Resolution of Degraded and Nonconforming Conditions, was issued for information in GL 91-18, on November 7,1991.
This guidance provided a process for licensees to develop a basis to continue operation or to piace the plant in a safe tudition and to take prompt conective action. It contained a number of provisions that relate to the role of 10 CFR 50.59 and the basis for continued operation of a facility.
Section 4.3.2, ' Changing the Current Ucensing Basis To Satisfy ari Appendix B Corrective Action,' stated:
A licensee may change the design of its plant es described in the FSAR in accordar.ce with 10 CFR 50.59, at any time. Whenever such changes are sufficient to resolve a degraded or nonconforming condition involving an SSC [ system, structure, or component]
that is subject both to Appendix B and 50.59, they may be used in lieu of restoring the affected equipment to its original
 
GL 9118, Revision 1 October 8,1997 I
Page 2 of 5 design. However, whenever such a change involves a unreviewed safety question (US0) or change in a technical specification (TS), the licensee must obtain a license amendment in accordance with 10 CFR 50.90 prior to oporuting (emphasis added) the plant with the degraded or nonconforming condition...
Section 4.5.1, " Justification for Continued Operation (JCO) Background,' stated:
The license authorizes the licensee to operate the plant in accordance with the regulations, license conditions, and the TS. If an SSC is degraded or nonconforming but operable, the license provides authorization to operate and the licensee does not need further justification. The licensee must, however, promptly identify and correct the condition adverse to safety or quality in accordance with 10 CFR Part 50, Appendix B, Criterion XVI.
A footnote to the flow chart attached to the Part 9900 guidance stated:
50.59 may be used to make a change in a facility, as described in the SAR, which w. m resolve the condition adverse to safety or quality so that the degraded and nonconforming condition no longer exists. Delay or partial correction of conditions adverse to safety or quality is considered a change in facility or procedures and subject to 50.59 review.
The NRC Inspection Manual Part 9900 guidance, '10 CFR 50.59 - Interim Guidance on the Requirements Related to Changes to Facilities, Procedures, and Tests (or Experiments)," issued in April 1996L specifically refers to the Pact 9900 attached to GL 91 18 for guidance conceming 10 CFR 50.59 in the resolution of degraded and nonconforming conditions.
As part of its roevaluation of the 10 CFR 50.50 process, the staff recognized that the guidance in GL 91-18 was not complete, and may in some respects be iaconsistent. Therefore, the staff developed additional guidance un the application of 10 CFR 50.59 to the resolution of degraded and nonconforming conditions.
The staffs proposed guidance was published for public comment, as part of draft NUREG-1606,
  ' Proposed Regulatory Guidance Related to implementatim of 10 CFR 50.59 (Changes Tests, or Experiments)," on May 7,1997 (62 FR 24947).
Descriotion of Circumstances The proposed guidance published for comment on May 7,1997, discussed the application of 10 CFR 50.59 to implementation of compensatory measures, how ' delay" should be interpreted, and how the guidance about ottaining a license amendment operating the facility with a condition involving a USQ should be interpreted. In this proposed guidance, the staff stated that implementation of compensatory measures reqnired a 10 CFR 50.59 evaluation with respect to the condition described in the final safety analysis report (FSAR) and that the staff would corisider delay to have occurred when a licensee has not implemented corrective action at the first available opportunity (considering need ft: analysis or parts, or the need to be in cold shutdown to complete the action), in any event not to exceed the next refwling outage. Finafiy, the staff proposed that when a licensee determined that resolution of a
 
i  .
GL 9118, Revision 1 October 8,1997 Page 3 of 5 nonconforming condition involved a USQ, the license amendment should be issued before the plant resumed operation from any shutdown (the NRC would not require a plant to shut down in such circumstances provided that SSCs required for operation were operable). Over the last several months, a number of nonconforming conditions have been identified at operating plants through licensee reviews and NRC inspections. Based on staff experience in dealir'g with these situations, the staff has concluded that a revision to the Part 9900 guidance," Resolution of Degraded and Nonconformind anditions,' was appropriate.
Many of the comments received in response to the FederalRegisternotice stated that the position that should be applied is more consistent with the discussion in Section 4.5.1 of the existing Part 9900 guidance, that is, if SSCs are operable but degraded, the license provides authority for continued operation, and existence of a USQ, by itself, should not be an impediment to a plant's ability to resume operation.
Commenters noted that the policy of not requiring plant shutdown but preventing plant restart was arbitrary, and had no basis in safety. Commenters also suggested that delay in implementation of corrective action is a matter for enforcement of 10 CFR Part 50, Appendix B, and not for requiring a 10 CFR 50.59 evaluation. The commenters also stated that requiring a 10 CFR 50.59 evaluation of compensatory measures against the condition described in the safety analysis report (SAR) would essentially preclude licensee implementation of compensating actions that enhance safety when degraded or nonconforming conditions are found.
On the basis of the staffs continuing review of the issues associated with nonconforming conditions and with interpretations of 10 CFR 50.59 requirements, and of the public comments that were received in response to the FederalRegister notice, the staff determined that it would be beneficial at this time to issue a revision to this inspection Manual Chapter 9900 guidance, even before other aspects of potential guidance are resolved, because of the impacts on plant operation. Therefore, through this generic letter, the NRC is notifying addressees of the issuance of the attached NRC Inspection Manual guidance.
Discussion As discussed in more detail in the attached guidance, the staff now concludes that the need to obtain NRC approval for the final resolution of a degraded or nonconforming condition does not affect the licensee's authority to continue operation (or restart from a shutdown), provided that necessary equipment is operable and the degraded equipment is not in conflict with any technical specification. Thus, Section 4.3.2 has been revised, and other conforming changes made, to note this change in staff guidance.
On July 21,1997, the Nuclear Energy Institute (NEI) submitted to the NRC a guidance document, NEl 95-07 (Final Draft), " Guidelines for 10 CFR 50.59 Safety Evaluations? Part of this guidance relates to applicability of 10 CFR 50.59 to degraded and nonconforming conditions.
 
GL 9118, Revision 1 October 8,1997 Page 4 of 5 Ths specific guidance is:
In the case of a nonconforming condition, there are three potential scenarios for addressing the condition:                                                                                                -
If the condition is accepted "as-is' resulting in something different than described in the SAR or is modified to something different than described in the SAR, then the condition should be considered a change and subjected to a 10 CFR 50.59 safety evaluation unless another regulation apnlies (i.e.,10 CFR 50.55a).
if the licensee intends to restore the SSC back to its p.                                    mndition (es described in the SAR), then this corrective action should oe performed in accordance with 10 CFR Part 50, Appendix B (i.e., in a timely manner commensurate with safety), and a 10 CFR 50.59 safety evaluation is not required.
If an interim compensatory action is taken to address the condition and involves a procedure change or temporary modification, a 10 CFR 50.59 review should be conducted and may result in a safety evaluation. The intent is to determine whether the compensatory action itself (not the degraded condition) impacts other aspects of the facility described in the SAR.
The staff finds this industry guidance acceptable with respect to the need for a 10 CFR 50.59 safety evaluation for degraded and nonconforming conditions. Therefore, the revised Part 9900 inspection Manual guidance references this industry guidance.
As noted in the Part 9900 guidance, the NRC will take enforcement action if it detennines that licensee                                              ,
corrective action (which may include submittal of a license amendment request) is not prompt, or that operabihty determinations are not sound. Enforcement action may also be taken for the circumstances that led to the existence of the degraded o nonconforming condition.
 
  .    -o GL 9118. Revision 1 October 8,1997 Page 5 of 5 This generic letter was not published for public comment because the issues covered by the revision were previously published for public comment in May 1997, and the staffs guidance is responsive to the comments received. This generic letter requires no specific action or response. If you have any questions about this matter, please contact the technical contact listed below,                          '
signed by Jack W, Roe, Acting Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contact: Eileen M. McKenna, NRR                                          -
301-415-2189 Email: emm@nrc. gov Attachments:                      _
1, inspection Manual Part 9900 Guidance, ' Resolution of Degraded and Nonconforming Conditions"
: 2. List of Recontly issued NRC Generic Letters
 
Attachment 1 GL 9118, Revision 1 October 8,1997 Page 1 of14 NRC INSPECTION MANUAL                              OTSB PART 9900: TECHNICAL GUIDANCE                            -
STS300EG.TG RESOLUTION OF DEGRADED AND NONCONFORMING CONDITIONS 9
e
 
Attachment 1 GL 91-18 Revision 1 October 8,1997 Page 2 of 14 RESOLUTION OF DEGRADED AND NONCONFORMING CONDITIONS Table of Contents
                                                                                                                                                                .P,jgg 1.0.
PU RPO S E AN D S COPE . .. . . . . . . . . . .. . . .. .. . . . . . . . . . . . . . .. . .. . . . . . . . . . . , . . . . . .. .. . . . .
2.0 DEFINITIONS...........................................,,,,.............................................,..........................2 2.1 C u rren t Ucensin g B asis .... ........... ... .... , .. . . . . .. .. . . . . .. . .... .. ... . .. .. ... . . .. . ... . . .. .. ..... ...
2.2 De sig n B asis . . .. . .. . . ... . . . .. . . . . . . . . . . .. . . . . . . . . . . . . . . . .. . .. . . . . , . . . . . .. . . . . . . . .. . . .
2.3 Degraded Condition . , ......... ............... ...... .......... .. . . .. ............ .. .. ............ ... ... ...... .. .. ..... .. . .,2 2.4 Nonconforming Condition . . . ....... . . ... . ..... .... . .. ...... ... ....... ... ,, .. . ... ... ...... ... . .... .. . . . ..... .. . .
2.5 Full Qu alification ....... .. . . . . . . . .. .,.. . ..... .. . .... .. ... ..... . . . ... . ... . ... .... ... .... .. . ...... . . . .. .. .. . . .
 
==3.0 BACKGROUND==
4.0 DISCU S SION OF NOTAB LE PROVISION S ... ..........,.. ......., ........, .... .... .... ... ........., ........ ...... . 3 4.1 Public Health an d S afety .... .. .. ...... .. . ..... .. . ........ .......... .. ... .... .. ... . .. . . . . ... . . ... . ... ... . . .. .. .
4.2 Operability Determinations ........... ....... ........ ..... ........... ............. .... ...... .. ..... .... . . . .... .... 3 4.3 The Current Licensing Basis and 10 CFR 50 Appendix B................., .......... ,, . . ... ........ ... 4 4.4 Discovery of an Existing But Previously Unanalyzed n    n or n...................................................................................................4 4,5 Justification for Continued Operation (JCO). .. ,, ,, .... .. . ........... ........... . . .. .,............4 4.5.1      B ackg rou nd . . . . . . . . . . . . . . . . .. . . . . . . . . .. . . . . . . . . . ... . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .
4.5.2      JCO Definition..      . . . . . . . .
                                                                                              ..................................                        .........5 4.5.3 Items for Consideration in a JCO...            . . . .  , , . . . .        . . .    ............5 4.5.4      Discussion of Industry-Type JCOs.. . ...                  . . . . . . . . . . . . . . .  ..............6
 
                                                                                                                          .            o-Attachment 1 GL 9118, Revision 1 October 8,1997 Page 3 of 14 4.6 Reasonable Assurance of Salsty .................................. . .................. ............. ........... .......... 6 4.7 _ Evaluation of Compensatory Measures. ,.......................................... ............. .................
4.8 Fin al Corrective Action .. ........... .. ... ...... ...... ... ..... ....... ... .. .... .. . ..... ........ ....          REFERENCE.........................................................................................................................,.9-9900 Degraded Conditions                          - ii -                                            Issue Date: 10/08/97
 
O
* l Attachment 1 GL 9118, Revision 1 October 8,1997 Page 4 of 14 RESOLUTION OF DEGRADED AND NONCONFORMING CONDITIONS 1.0      PURPOSE AND SCOPE:
To provide guidance to NRC inspectors on resolution of degraded and nonconforming conditions affecting the following systems, structures, or components (SSCs):
(i)      Safety-related SSCs, which are those relied upon to remain functional during and following design baris events (A) to ensure the integrity of the reactor coolant pressure boundary, (B) to ensure the capability to shut down the reactor and maintain it in a safe shutdown condition, or (C) to ensure the capability to prevent or mitigate the consequences of accidents that could result in potential offsite consequences comparable to the 10 CFR Part 100 guidelines. Design basis events are defined the same as in 10 CFR 50.49(b)(1).
(ii)      All SSCs whose failure could rrevent satisfactory accomplishment of any of the reqtired functions identified in (i) A, B, and C.
(iii)    All SSCs relied on in the saf3ty analyses or plant evaluations that are a part of the plant's current licensing basis. Such analyses and evaluations include those submitted to support license amendment requests, exemption requests, or relief r2 quests, and those submitted to demonstrate comphance with the Commission's regulations such as fire protection (10 CFR 50.48),
environmental qualification (10 CFR 50.49), pressurized thermal shock (10 CFR 50.61),
anticioated transients without scram (10 CFR 50.62), and station blackout (10 CFR 50.63).
(iv)      Any SSCs subject to 10 CFR Part 50, Appendix B.
(v)      Any SSCs subject to 10 CFR Part 50, Appendix A, Criterion 1.
(vi)      Any SSCs explicitly subject to facility Technical Specifications (TS).
(vii)    Any SSCs subject to facility TS through the definition of operability (i.e., support SSCs outside TS).
(viii) Any SSCs described in tne final safety analysis report (FSAR).
This guidance is directed toward NRC inspectors who are reviewing actions of licensees that hold an operating license. Although this guidance generally reflects existing staff practices, Issue Date: 10/08/97                                                              9900 Degraded Conditions
________A
 
                                                                                                                . o Attachment 1 GL 91-18, Revision 1 October 8,1997 Page 5 of 14 application to specific plants may constitute a backfit. Consequently, significant differences in licensee l  practices should be discussed with NRC management to ensure that the guidance is applied in a i
reasonabka and consistent raanner for all licensees.
l 2.0      DEFINITl0tjS                                                                                        '
(
2.1      Current Licensino Basis Current licensing basis (CLB)is the set of NRC requirements applicable to a specific plant, and a licensee's wntten commitments for assuring compliance with and operation within applicable NRC requirements and the plant specific design basis (including all modifications and additions to such commitments over the life of the license) that are docketed and in effect. The CLB includes the NRC regulations contained in 10 CFR Parts 2,19,20,21,30,40,50,51,55,72,73,100 and appendices thereto; orders; heense conditions; exemptions, and TS. It also includes the plarit-specific design basis information defined in 10 CFR 50.2 as documented in the most recent FSAR as required by 10 CFR 50.71 and the hcensee's commitments remaining in effect that were made in docketed licensing correspondence such as licensee responses to NRC bulletins, generic letters, and enforcement actions, as well as licensee commitments documented in NRC safety evaluations or licensee event reports.
2.2    Desion Basis Design basis is that body of plant-specific design bases information defrned by 10 CFR 50.2.
2.3    Deoraded Condition A conditior, of an SSC in which there has been any loss of quality or functional capability.
2.4    Nonconformino Condition A condition of an SSC in which there is failure to meet requirements or licensee commitments. Some examples of nonconforming conditions include the following:
1.
There is failure to conform to one or more applicable codes or standards specified in the FSAR.
          ?
As-built equipment, or as-modified equipment, does not meet FSAR descriptions.
3.
Operating experience or engineering reviews demonstrate a design inadequacy.
: 4.            Documentation required by NRC requirements such as 10 CFR 50.49 is not available or deficient.
9900 Degraded Conditions                                                                Issue Date: 10/08/97
 
o .
Attachment 1 GL 9118, Revision 1 October 8,1997 Page 6 of 14 2.5    Fuu Oualification
                                                                                      ~
Full qualification constitutes conforming to all aspects of the current licensing basis, including codes and standards, design criteria, and commitments.
 
==3.0    BACKGROUND==
A nuclear power plant's SSCs are designed to meet NRC requirements, satisfy the current licensing bas and conform to specifed codes and standards. For degraded or nonconforming conditions of these SSCs, the licensee may be required to take actions required by the TS. The provisions of Title 10 of the ' Code of Federal Regulations" (10 CFR), Part 50, Appendix B, Criteria XVI, may apply requiring the licensee to identify promptly and correct conditions adverse to safety or quality. Reportirig may be required in accordance with Sections 50.72,50.73, and 50.9(b) of 10 CFR Part 50,10 CFR Part 21, and the TS.
Collectively, these requirements may be viewed as a process for licensees to develop a basis to continue operati'q or to place the plant in a safe condition, and to take prompt corrective action. Changes to the -
facihty in accordance with 10 CFR 50.59 may be inade as part of the corrective action required by Appendix B. The process displayed by means of the attached chart titled, ' Resolution of Degraded and Nonconforming Conditions," recognizes these and other provisions that a licensee may follow to restore or establish acceptable conditions. These provisions are success paths that enable licensees to continue safe operation of their facilities.
4.0 DISCUSSION OF NOTABLE PROVISIONS 4.1 Pub!ic Health and Safety All success paths, whether specifically stated'or not, are first directed to ensuring public health and safety and second to reswring the SSCs to the current licensing basis of the plant as an acceptable level of safety. Identification of a degraded or nonconforming condition that may pose an immediate threat to the public health and safety requires the plant to be placed in a safe condition.
Technical Specifications (TS) address the safety systems and provide Limiting Conditions for Operation (LCOs) and Allowed Outage Times (AOTs) required to ensure public health and safety.
4.2 Operability Determinations For guidance on operability see the Inspection Manual, Part 9900, ' OPERABLE / OPERABILITY:
ENSURING THE FUNCTIONAL CAPABILITY OF A SYSTEM OR COMPONENT,' and see the inspection Manual Part 9900,' STANDARD TECHNICAL SPECIFICATIONS STS SECTION 1, OPERABILITY."
issue Date: 10/08/97                                                          9900 Degraded Conditions
 
Attachment 1 GL 91 18, Revision 1 October 8,1997 Page 7 of 14
                                                                                                      ~
4.3 The Current Licensino Basis and 10 CFR 50. Aooendix B The design and operation of a nuclear plant is to be consistent with the current licensing basis. Whenever degraded or nonconforming conditions of SSCs subject to Appendix B are identified, Appendix B requires prompt corrective action to correct or resolve the condition. The licensee must ecWsh a time frame for completion of corrective action. The timeliness of this corrective action should be commensurate with the safety significance of the issue. The time frame gcveming corrective action begins with the discovery of the condition, not with the time when it is reported to the NRC. In determining whether the licensee is making reasonable efforts to complete corrective action promptly, NRC will consider whether corrective action was taken at the first opportunity, as determined by safety significance (effects on operability, significance of degradation) and by what is necessary to implement the corrective acion. Factors that might be included are the amount of time required for design, review, approval, or procurement of the repairlmodification; availability of specialized equipment to perform the repair; otthe need to be in a hot or cold shutdown to implement the actions. The NRC expects time frames longer than the next refueling outage to be explicitly justified by the licensee as part of the deficiency tracking documentation. If the licensee does not resolve the degraded or nonconforming condition at the first available opportunity or does not appropriately justify a longer completion schedule, the staff would conclude that corrective action has not been timely and would consider taking enforcement action.
4.4              Discovery of an Existino But Previousiv Unanalyzed Condition or Accident in the course of its activities, the licensee may discover a previously unanalyzed condition or accident.
Upon discovery of an existing but previously unanalyzed condition that significantly compromises plant safety, the licensee shall report that condition in accordance with 10 CFR 50.72 and 50.73, and put the plant in a safe condition.
For a previously unanalyzed condition or accident that is considered a significant safety concem, but is not part of the design basis, the licensee may subsequently be required to take additional action after consideration of backfit issues (see $ection 50.109(a)(5)).
4.5              Justification for Continued Operation (JCO) 4.5.1              Background The license authorizes the licensee to operate the plant in accordance with the regulations, license conditions, and the TS. If an SSC is degraded or nonconforming but operable, the license establishes an acceptable basis to continue to operate and the licensee does not need to take any further actions. The licensee must, however, promptly identify and correct the condition adverse to safety or quality in accordance with 10 CFR Part 50, Appendix B, Criterion XVI.
9900 Degraded Conditions                                                                            Issue Date: 10/08/97
 
s O Attachment 1 GL 9118, Revision 1 October 8,1997 Page 8 of 14 The basis for this authority to continue to operate arises because the TS contain~the specific characteristics and conditions of operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immedid.e threat to public health and safety. Thus, if the TS are satisfied, and required equipment is operable, and the licensee is correcting the degraded or nonconforming condition in a timely manner, continued plant operation does not pose an undue risk to public health and safety.
Under certain defined and limited circumstances, the licensee may find that strict compliance with the TS would cause an unnecessary plant action not in the best interest of public health and safety. NRC review and action is required prior to the licensee taking actions that are contrary to compliance with the license conditions or TS unless an emergency situation is present such that 10 CFR E0.54(x) and (y) is applied. A JCO, as defined herein for general NRC purposes, is the licensee's technical basis for requesting NRC responses to such action.
4 4.5.2                              JCO Definition A Justification for Continued Operation 1 (JCO) is the licensee's technical basis for requesting authorization to operate in a manner that is prohibited (e.g., outside TS or license) absent such authorization. The preparation of JCOs does not constitute authorization to continue operation.
l    4.5.3                              Items for Considerauon in a JCO Some items which are appropriate for consideration in a licensee's development of a JCO include:
Availability of redundant or backup equipment Compensatory measures including limited administrative controls Safety function and events protected against Conservatism and margins, and Prcbability of needing the safety function.
Probabilistic Risk Assessment (PRA) or Individual Plant Evaluation (IPE) results that determine how operating the facility in the manner proposed in the JCO will impat.t the core damage frequency.
        ' Regulations, generic letters, and bulletins may provide direction on specific iscue JCOs, which do not require that they be submitted. Licensees may a!so use the JCO for situations other than for operating in a prohibited manner. The JCO term has been used in Generic Letters 88-07 on Environmental Qualifications of Electrical Equipment and 87-02 on Seismic Adequacy. Licensees should continue to follow earlier guidance regarding the preparation of JCOs on specifc issues.
Issue Date: 10/08/97                                                                            9900 Degraded Conditions I
                                                                                                                      - _ _ _            \
 
O  a Attachmenti GL 9118, Revision 1 October 8,1997 Page 9 of 14 4.5.4          Discussion ofIndustry Type JCOs Currently, some licensees refer to two other documents or processes as JCOs that are not equivalent to and do not perform the same function as the NRC-recogr*ed JC0 (as defined in 4.5.2). This is an acceptable industry practice and to the extent the industry sCO fulfills other NRC requirements, the JCOs l  will be selectively reviewed and audited accordingly.
1 In the first industry-type JCO, the licensee may consider the entire process depicted in the attached chart as a single JCO that inciudes such things as the basis for operability, PRA, corrective action elements, and attemative operations.
In the second industry type JCO, the licensee may consider the documentation that is developed to support facility operation after the operability decision has been made as a JCO. This documentation can cover any or all of the items listed under ' interim Operation' on the attached chart.
Although the 'JCO' is used d"erently by some licensees, the NRC concem is that the operability decision is correct, documentation of licensee's actions are appropriate, and submittals to the NRC are complete.
The licensee's documentation of the JCO is normally proceduralized through the existing plant record system, which is auditable.
4.6      Reasonable Assurance of Safety For SSCs that are not expressly subject to TS and that are determined to be inoperable, the licensee should assess the reasonable assurance of safety, if the assessment is successful, then the facility may continue to operate while prompt corrective action is taken. Items to be considered for such an assessmentinclude the following:
Availability of redundant or backup equipment Compensatory measures including limited administrative controls Safety function and events protected against Conservatism and margins, and Probability of needing the safety function.
PRA or Individual Plant Evaluation (IPE) results that determine how operating the facility in the manner proposed in the JCO will impact the core damage frequency.
4.7      Evaluation of Comoensatory Measures in its evaluation of the impact of a degraded or nonconforming condition on plant operation and on operability of SSCs, a licensee may decide to implement a compensatory measure as an interim step to restore operability or to otherwise enhance the capability of SSCs until the final corrective action is complete. Reliance on a compensatory measure for operability should be an important consideration in establishing the ' reasonable time frame
* to complete the corrective action process. NRC would normally expect that conditions that require interim compensatory measures to demonstrate operability would be resolved more promptly than conditions that are not dependent on compensatory measures to show operability, because 9900 Degraded Conditions                                                                            Issue Date: 10/08/97
 
Attachment 1
)                                                                                  GL 9118, Revision 1 October 8,1997 Page 10 of 14 such reliance suggests a greater degree of degradation. Similarly, if an operability determination is based upon operator action, NRC would expect the nonconforming condition to be resolved expeditiously.
On July 21,1997, the Nuclear Energy Institute (NEI) submitted to the NRC a guidance document, NEl 96-07 [ Final Draft), " Guidelines for 10 CFR 50.59 Safety Evaluations ' Part of this guidance relates to applicabihty of 10 CFR 50.59 to degraded and nonconforming conditions. With respect to the use of    .
c,ompensatory measures, the guidance states:
If an interim compensatory action is taken to address the condition and involves a procedure change or temporary modification, a 10 CFR 50.59 review should be conducted and may result in a safety evaluation. The intent is to determine whether the compensatory action itself (not the degraded condition) impacts other aspects of the facility descrioed in the SAR.
The staff concludes that this is an acceptable approach for dealing with compensatory actions within the context of a corrective action process.                                        .
In considenng whether a compensatory measure may affect other aspects of the facility, a licensee should pay particular attention to ancillary aspects of the compensatory measure that may result from actions taken to directly compensate for the degraded condition. As an example, suppose a licensee plans to close a valvo to isolate a leak. Although that action would temporarily resolve the leak, it has the potentia to affect flow distribution to other components or systems, may complicate required operator responses,        7 or could have other effects that should be evaluated before the compensatory measures are implemented, in j
accordance with 10 CFR 50.59, should the evaluation determine that implementation of the compensato action itself would involve a TS change or an unreviewed safety question (US0), NRC approval, in accordance with 10 CFR 50.90 and 50.92, is required prior to implementation of the compensatory action.
4.8      finalCorrective Action The responsibility for corrective action rests squarely on the licensee. A licensee's range of corrective Ection could include (1) full restoration to the SAR-described condition, (2) NRC approval for a change to its licensing basis to accept the as-found condition as is, or (3) some modification of the facility other than restoration to the original FSAR condition. If corrective action is taken so that the degraded or nonconforming condition is restored to its original configuration, no 10 CFR 50.59 evaluation is required.
The 10 CFR 50.59 process is entered when the final resolution to the degraded or nonconforming condition is to be different than the established FSAR requirement. At this point, the licensee is planning (in a prospective sense) to make a change to the facility or procedures as described in the SAR. The proposed change is now subject to the evaluation process established by 10 CFR 50.59. A change can be safe, but can still require NRC approval. The proposed final resolution can issue Date: 10/08/97                                                        9900 Degraded Conditions
 
                                                                                                                                                  . u 1
Attachment 1 l-                                                                                                                  GL 9118, Revision 1 l
October 8,1997 Page 11 of 14 be under staff review and not affect the continued operation of the plant, because interim operation is being governed by the processes of the operability determination and corrective action of Appendix B.
l-In two situations, the identification of a final resolution or final corrective action would trigger a 10 CFR 50.59 evaluation, unless another regulation applies (i.e.,10 CFR 50.55a): (1) when a licensee decides to change its facility or procedures to something other than full restoration to the FSAR-described condition, as the final corrective action, or (2) when a licensee decides to change its licensing basis as described in the SAR to accept the degraded or nonconforming condition as its revised licensing basis. This guidance is consistent with the July 21,1997, revision of NEl 96-07.
Change to Facility or Procedures The first circumstance is if the licensee plans for its final resolution of the degraded or nonconforming condition to include other change (s) to the facility or procedures in order to cope with the (uncorrected, including only partially corrected) nonconforming condition. Rather than fully correcting the nonconforming condition, the licensee decides to restore capability or margin by another change, in this case, the licensee needs to evaluate the change from the SAR<lescribed condition to the final condition in which tne licensee proposes to operate its facility. If the 10 CFR 50.59 evaluation concludes that a change to the TS or a USQ is involved, a license amendment must be requested, and the corrective action process is not complete until the approval is received, or other resolution occurs.
Change to Current Licensing Basis The other situation is a final resolution in which the licensee proposes to change the current licensing basis to accept the as-found nonconforming condition, in this case, the 10 CFR 50.59 evaluation is of the change from the SAR-described condition to the existing condition in which the licensee planc to remain (i.e., the licensee wil! exit the corrective action process by revising its licensing basis to document acceptance of the condition). If the 10 CFR 50.59 evaluation concludes that a change to the TS or a USO is involved, a license amendment must be requested, and the corrective action process is not complete until the approval is received, or other resolution occurs. In order to resolve the degraded or nonconforming condition without restoring the affected equipment to its original design, a licensee may need to obtain an exemption from 10 CFR Part 50 in accordance with 10 CFR 50.12, or rehef from a design code in accordance with 10 CFR 50.55a.
The use of 10 CFR 50.59,50.12, or 50.55a in fulfillment of Appendix B corrective action requirements does not reheve the licensee of the responsibility to determine the root cause, to examine other affected systems, or to report the original condition, as appropriate.
9900 Degraded Conditioris                                                            Issue Date: 10/08/97
 
-                                                                                                                            Attachment 1 GL 9118, Revision 1
}                                                                                                                            October 8,1997 Page 12of 14 in both of these situations, the need to obtain NRC approval for a change (e.g., because it involves does not affect the licensee's authority to operate the plant. The licensee may rpake mode changes, restart from outages, etc., provided that necessary equipment is operable and the degraded conditicCs not 19 conflict with the TS or the license. The basis for this position was previously discussed in Section 4.5.1.                                                                                                                          '
ENFORCEMENT
'                          ff the licensee, without good cause, does not correct the nonconformance at the first available opportu the staff concludes that the licensee has failed to take prompt corrective action and, thus, is in violation of 10 CFR Part 50 Appendix B (Criterion XVi).2 When the NRC concludes that corrective action to implemen the final resolution of the degraded or nonconforming condition is not prompt, or that the operability determination is not valid, enforcement action (Notice of Violation, orders) will be taken. Enforcement action may include restrictions on continued operation.                                                                          I implementation of complete corrective action within a reasonable time frame does not mitigate the potential for taking enforcement action for the root causes that initially created the degraded or nonconforming condition or for violations of other regulatory requirements. The nonconforming condition may have resulted from (1) earlier changes performed without a 10 CFR 50.59 evaluation or (2) inadequate reviews; or may be a de facto change for which the facility never met the SAR description. The l                        staff may determine that the ' change' from the FSAR-described condition to the discovered nonconinmiing condition involved a USQ (or a TS change), and that enforcement action is appropriate for the time frame up to time of discovery.
5.0      ft1FERENCE Sea attached charts titled, ' Resolution of Degraded and Nonconforming Conditions.'
END 2Since Appendix B is only applicable to safety-ralated SSCs, this approach could not be used if the delay in resolution of a nonconforming condition from the SAR involved only non-safety-related SSCs and did not affect any safety related SSCs. However, NRC expects licensees to take corrective action for nonconformances with the SAR consistent witn Criterion XVIin a time frame commensurate with safety.                                -
Issue Date: 10/08/97                                              9-                                      9900 Degraded Conditions
 
A        t pa nas y                  #*,
UNITED STATES                              *
:=                                NUCLEAR REGULATORY COMMISSION WASHINGTON D.C. 30MH001
                  . ,,.                                              September 5, 1997 MEETING PARTICIPANT:            NUCLEAR ENERGY INSTITUTE          ,
 
==SUBJECT:==
 
==SUMMARY==
OF APRIL 30. 1997. MEETING WITH THE NUCLEAR ENERGY INSTITUTE (NEI) REGARDING GENERIC LETTER (GL) 96 06. " ASSURANCE OF E0VIPMENT OPERABILITY AND CONTAINHENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS" Or, April 30, 1997, represent 6tives of NE! met with representatives of the Nuclear Regulatory Comission (NRC) at the NRC's offices in Pockville.
Maryland. Attachment 1 provides a list of meeting attendees.
The meeting was requested by NEI representatives to discuss issues pertaining to GL 96-06. " Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions." Prior to the meeting NEI sent NRC a proposed agenda and a list of questions that it wanted the staff to address during the meeting. The staff's response to these questions is summarized in this meeting summary. The agenda is contained in Attachment 2.
The NRC staff began the meeting by making a presentation that included an overview of the GL, a summary of the industry responses to the GL received by the NRC to date, and a brief discussion on each of the technical . issues in the      '
GL, Attachment 3 contains the overhead slides that were used during the NRC presentation. The NRC staff stated that it has been disappointed with the 120 day responses.to the GL because of some licensees' questionable schedules for resolvir.J the problems identified in the GL, Also, some of the responses submitted were incomplete. The staff has been contacting individual licensees in an attempt to obtain a better understanding of licensee responses and schedules. An industry representative stated that in order to plan outaget modification packages must be completed at least 6 months prior to the beginning of an outage: therefore, if the GL 96 06 modifications cannot be installeo during the upcoming outages, the modifications will not be installed until 18 to 24 months later (a refuel cycle), which may put implementation of some plant modifications into the year 1999. The staff stated that licensees became aware of the issues on September 30. 1996. when the GL was issued and-guidance in GL 91-18 ("Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability") suggests that degraded systems should be restored to within the original criteria during the next refueling outage.
The NEI representatives next made their formal presentation which included a restatement and clarification of the questions that were given to the NRC to address during the meeting. The NEI presentation materials are provided as Attachment 4, which includes the formal questions given to the NRC prior to the meeting. The NRC staff addressed NEI's questions as each was presented.
CONTACT:      Beth Wetzel. NRR 301-415 1355 3
                      $YOllOONU
 
e      s ll 1
2-                September 5, 1997 Individual answers to each NEI question have not been provided in this meeting                -
summary since the staff believas the discussions in this summary and its attachments cover the basic issues of the' questions.                    ,
i During the meeting, the staff discussed some of the issues associated with -
waterhammer and two phase flow.          In general, plants were licensed assuming single phase flow conditions in the service water and component cooling water systems. The staff expects that licensees will make modifications as              -
necessary to conform to the original system design basis, or seek NRC review and approval for any proposed changes to the design basis as appropriate. The staff also cautioned that some system modifications (e.g., changing a closed loop cooling water system from one that is vented to one that is pressurized) may introduce new vulnerabilities or warrant new Technical Specification requirements, and prior NRC review and ap3roval may be required.
Finally, the staff emphasized that evaluations of waterlamer and two phase
!                            flow conditions must include worst case design basis assumptions, such as l
single failure considerations, i
The staff discussed the use of American Society of Mechanical Engineers (ASME)
Section Ill Appendix F criteria for the evaluation of thermal overpressure events. Ap)endix F provides rules for the evaluation of components for ASME Code Level ] service loadings. These service loadings are typically specified for plant falted design conditions, The staff has endorsed the use of the Appendix F criteria for operability evaluations in GL 91-18.
Appendix F contains criteria for inelastic analysis of components. Licensees may be able to demonstrate the long-term acceptability of some piping runs between containment isolation valves for the thermal overpressure event using the inelastic criteria of Appendix F. If licensees decide to use Appendix F i
                          -criteria by adding it to their current licensing basis and they determine that use of the criteria creates an unreviewed safety question through a 10 CFR 50.59 evaluation.- then NRC prior review and approval would be required.
'                          The staff notes that the Appendix F criteria apply only to load combinations associated with plant faulted conditions. Licensees would have to describe                      ;
the applicable load combinations used in the Appendix F evaluatson and the basis for determining the applicable load combination associated with the thermal overpressure event.
Another topic of discussion involved the appropriate set point pressure fcr thermal relief valves. The staff stated tlat the set point pressure should be determined such that the applicable design criteria are not exceeded.
Representatives from the Electric Power Research Institute (EPRI) made a presentation on two projects (one for the overpressurization of piping and one l                          for waterhamer) that it is overseeing to support the industry in resolving the technical issues identified in GL 96 06. The EPRI presentation is provided in Attachment 5.
NEI and the staff discussed the use of the EPRI test results to verify the adequacy of piping for thermal overpressure events.
The staff expressed its view that it Nuld be difficult to design a bounding test configuration: however, the use of testing on piping configurations that l-
 
  ~                                            - . _ _ _ _ _                            __ _._ _ _ _ _ _ . _ _ .
        ,          e i
i
                                                                                                          ~
3*                            September 5, 1997 are representative of some common plant piping layouts might be useful.                                                                      NEI
'                        and the staff also discussed the use of testing to verify an analytical procedure that could be used to evaluate unique plant configurations.                                                                    The staff considers the use of such an approach to be reasonable.
During the meeting, there was considerable discussion regarding 10 CFR 50.59 and US0s (unreviewed safety questions) with respect to GL 96 06, The industry is concerned with the GL- 96 06 issues and whether the staff considers                                                                        '
potential resolutions of GL 96 06 issues to involve US0s or not. particularly with respect to the use of ASME Section III Appendix F criteria for piping                                                                                      !
lo20s for other than interim operability evaluations. Indust"y                                                                                                !
representatives went as far as to sa                                                                                                                          '
criteria was determined to be a US0,ythey                                              thatwould if long-term installuse of Appendix relief valves forF the piping susceptible to overpressurization. then request use of the Appendix F criteria and remove the relief valves after use of Appendix F was granted.                                                                                      '
NRC policy has been that if a plant has a USO and it shuts do#, the plant cannot restart until the US0 has been reviewed ar.d approved by the NRC,                                                                                        t However, some of the modifications associated with G. 9646 issues could take a considerable amount of time to design, procure, schedule, and implement,
"                      thereby being impractical- to implement prior to the next startup, The staff acknowledged the industry representatives' concerns with respect to US0s but was not able to answer their specific questions at the meeting because they were associated with policy decisions. The staff has been considering a possible revision to the existing policy which, among other changes. would eliminate the restriction on plant restart with a USO. provided that the                                      -
system, structure, or component in question is operable and the applicable regulations and the technical specification requirements are being met. A final determination on this policy issue has not been reached.
u          .
Beth A. Wetzel, P oject Manager Project Directorate 111 1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Attachments:                          1.          List of Attendees
: 2.          Meeting Agenda
: 3.          NRC Handouts 4          NEI Handouts
: 5.          EPRI Handouts 4
e,,  ,w-w-wwwr  =w    -.w-,-+------- = - * = . - - - - -    c,w-mw--    -e,-w..,    .-W4      w-.-c. -    -    -v e--,v                  v-~----rvm v,+-  -r-    e  , rw--ce
 
l tgIIING ATTENDEES
                                  !Bng                                                    Oraanization L. Marsh                                                      NRC                            -
G. Hubbard                                                    NRC G. Hammer                                                    NRC K. Manoly                                                    NRC R. Wessman                                                  NRC
: 3. Wetzel                                                    NRC R. Lobel                                                    NRC J. Fair                                                      NRC J. Tatum                                                    NRC C. Saadu                                                    NRC E, McKenna                                                  NRC S. Greco                                                    Wisconsin Electric T. Wroblewski                                                Wisconsin Electric S. Denny                                                    PECO Nuclear R. Raridels                                                  Comed R. Gamberg                                                  Duke Power Co.
D. Murphy                                                    Bechtel A. Singh                                                    EPRI B. Demars                                                    Virginia. Paver T. Sutter                                                    Bechtel E. May                                                      Virginia Power i
D. Stellfox                                                  Inside NRC P. Tamburro                                                  GPU Nuclear S. Mixon                                                    NUS l                                D. McGuigan                                                  Scientech l                                P. Okas                                                      NYPA W. Birely                                                    NYPA J. del Mazo                                                  Pacific Gas & Electric G. Davant                                                    Entergy T. Esselman                                                  Altrcn Corporation S. Gosselin                                                  EPRI t
L D. Modeen                                                    NEI J. Minichello                                                  Comed S. Plymaie                                                    GPU Nuclear G. Gyrkan                                                      GPU Nuclear ATTACHMENT 1                  .
i
 
                                                                                                                                                                      \
PROPOSED AGENDA NRC/NEI MzRTENG ONGL E06 APa!L 80.1997 mh-                                                                                      BasPONSIRLK                      i FAan
: 1.      Welcome and Opening Remaria                                                    NRC NEI                            ,
S.      GL 96-06 Submittala: Status and Observat!ans                                    NRC a
: 3.    -
GL E06: Questions and Responaos                                        .      NEI NRC
: 4.      EPRI GL E06 programs on Over Pressurization and                                EPRI Watarh==cr (Tentative)                                                      .
: 5. Concluding Remarks                                                              NRC NEI
: 6. Adjournment                                  ,
{
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1 i
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Attachment 2
 
MEETING WITH NEI REGARDING GENERIC LETTER 96-06 APRIL 30,1997 l
GL 96-06, " ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS" l
TECHNICAL BRANCHES: SPLB, EMEB AND SCSB LEAD PM: BETH WIC12EL 5
h                                                          ~
j                                                                                    .
2
 
i J
PRESENTATION i
  + OVERVIEW OF GENERIC LETTER - BETH WEIZEL
  + INDUSTRY RESPONSES - CORETTA SAADU i
  + CONTAINMENT FAN COOLER ISSUES - JIM TATUM l + THERMAL OVERPRESSURE - KAMAL MANOLY
; + NEI DISCUSSION QUESTIONS - ALL
                                      ~
2 l
 
I SCOPE OF GL 96-06
}  (1)  POTENTIAL WATERHAMMER IN THE COOLING l        WATER SYSTEMS SERVING THE CONTAINMENT AIR CO9LERS DURING A LOSS-OF-COOLANT ACCIDENT (LOCA) OR A MAIN STEAM LINE BREAK (MSLB) o
;          DIABLO CANYON o HADDEM NECK
;  (2)  POTENTIAL TWO-PHASE FLOW IN THE COOLMG                  '
WATER SYSTEMS SERVING THE CONTAINMENT AIR COOLERS DURING POSTULATED LOCA AND MSLB SCENARIOS                                                ,
O  FI'. BEACH i
o  SALEM L
(3)  THERMALLY INDUCED OVERPRESSURIZATION 'OF ISOLATED PIPING INSIDE CONTAINMENT O  BEAVER VALLEY o MAINE YANKEE 3
                                                              ]l
__i
 
  - ..  - -    --. . .  - - .    - . - - - . - -        -- = . ..  .- --
REASONS FOR ISE_UANCE OF GENERIC LErrE_R
      +      THE PROBLEMS IDENTIFIED IN THE LICITER ARE COMPLIANCE ISSUES
      +      THESE PROBLEMS WERE IDENTIFIED AT SEVERAL                        '
PLANTS AND ALL PLANTS ARE POTENTIALLY                            ,
SUSCEPTIBLE l
      +      TWO OPERATING PLANTS SHUT DOWN BECAUSE OF THE ISSUES AND HARITWARE MODIFICATIONS WERE 4
REQUIRED l
:      +      NRC WAS DEALING WITH THESE ISSUES ON A CASE-BY-CASE BASIS                            '
J 4
                                .                  ._.  -          . _ _ __ n
 
REQUESTED INFORMATION
                  +                                                    GL ISSUED SEPTEMBER 30,1996 l                                                                                                                                                            !
i
                  +                                                    30 DAY RESPONSES l                  +                                                    120 DAY RESPONSES DUE JANUARY 28,1997 1                                                                        o                          IF SUSCEPTIBLE                                          i o                          ACTIONS TAKEN 1
0                          BASIS FOR OPERABILITY o                          CORRECTIVE ACTIONS AND SCHEDULE 2
I
;                                                                                                                      5 2
l
 
4 1
)                                                                  -
REVIEW PLAN
;                                                                    i
!  +      TIMELINE                                                  i i
o SCREENING REVIEW o ISSUE MAJORITY OF SEs BY 18 MONTHS 0  CONSIDERING TI l
  +      REVIEW RESPONSIBILITIES i
;l        SPLB - CFCU ISSUES                                        '
I          EMEB - OVERPRESSURIZATION OF PIPING ISSUES SCSB - CONTAINMENT PARAMETERS 4
g            I i
6                                    I i                                                    __ _ _ _ _ _ _
 
i
^
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PRELIMINA_RY GL RESPONSE
 
==SUMMARY==
 
    +  7 PLANTS NOT SUSCEFFIBLE TO ANY OF THE ISSUES j  +- 2 PLANTS SUSCEFFIBLE ONLY TO WATERHAMMER AND TWO PHASE FLOW
    +  45 PLANTS SUSCEPTIBLE ONLY TO THERMAL OVERPRESSURIZATION i
,  4  51 PLANTS SUSCEFFIBLE TO BOTH CONTAINMENT FAN COOLER ISSUES AND THERMAL                                              j OVERPRESSURIZATION i
i I
1                                                      '
l                                  7                                            -
                                                    .. _      _      ___-_.-_I
 
OUTLIERS / STAFF FOLLOWUP ACTIONS
~
  + ADDITIONAL SUBMITTALS:
i o  JUNE SUBMITfAL - 6 PLANTS                                    i o  JULY SUBMITTAL - 7 PLANTS                                    i i
o  AUGUST SUBMITTAL - 3 PLANTS o  OCTOBER AND BEYOND - 5 PLANTS t
i i
  + MODIFICATION SCHEDULES EXTENDING INTO SPRING 1998 AND BEYOND: 7 PLANTS i
8 i
 
  . -.  ..        -    .  .  - -    _.  ..      . - - -          .=  --_-            _.
WATER HAMMER /TWO-PHASE FLOW ISSUES i
i      +        ANALYTICAL ACCEPTANCE OF TIIE CONDITION
!                O NEW ANALYSIS / METHODOLOGY /UNCERTAINrIES
}                o NOT PREVIOUSLY REVIEWED / ACCEPTED BY NRC
;      +        SYSTEM MODIFICATIONS I                o SATISFY DESIGN BASIS l                0 NEW VULNERABILITIES & FAILURE MODES j                o TS CONSIDERATIONS i
        +      SINGLE FAILURE /FMEA CONSIDERATIONS o WORST-CASE / BOUNDING CONDITIONS                                                              '
        +      POST-ACCIDENT USE OF NON-ESSENTIAL CFCs
        +      IMPACT ON PLANT OPERATION o
USQ DETERMINATION / LICENSE SUBMirrAL I              o COMPLEX REVIEW / LIMITED RESOURCES
                                                                      ~
l                                    9                                                                        -
e
 
THERMAL OVERPRESSURIZATION
    +  OVERPRESSURIZATION OF ISOLATED PIPING IS            l ADDRESSED IN ANSI B31.1 AND ASME DESIGN CODES
    +  AFFECTED PIPING RANGES FROM 3/8-INCH TO 20-INCH O.D.
    +  SOME MEMBERS OF THE ASME CODE DISAGREE THAT THE PHENOMENON IS A FAULTED DESIGN CONDITION
    + PROPOSED ASME CODE CHANGE TO DECLARE AFFECTED PIPING SYSTEMS MEET THE CODE WITHOUT FURTHER ANALYSIS FOR TEMPERATURE INCREASE UP TO 250 F 10 l
 
THERMAL OVERPRESSURIZATION (cont'd)
{'
; + STAFF DOES NOT ENDORSE PROPOSED CHANGE i  SINCE IT IS BASED ON SIMPLISTIC ANALYSIS OF IDEAL PIPING CONFIGURATION                          !
{
( + APPENDIX F TO SECTION III OF ASME CODE EMPLOYED IN SOME OPERABILITY DETERMINA-TIONS. LICENSE AMENDMENT MAY BE REQUIRED FOR LONG-TERM RESOLUTTON AT SOME FACILITIES 4 PROPOSED SOLUTIONS INCLUDE THERMAL RELIEF VALVE, ROY1 URE DISK WITII EXPANSION 1
CHAMBER, DRAINAGE OF LINE DURING NORMAL 4
OPERATION, PERFORM INELASTIC ANALYSIS OF PIPING (APPLICATION OF APPENDIX F), FAILURE IJAS NO SAFEIT CONSEQUENCES, AND DETAU;ED j  HEAT TRANSFER ANALYSIS TO DEMONSTRATE
!  THAT PRESSURE DOES NOT EXCEED ALLOWABLE            i l
                                        ~              '
}                      11                            -
 
I i
i i
                                                                                                                                                                              '~
1 i
i                                                                                                                                                                                                                        ;
INDUSTRY -- NRC STAFF MEETING l                                                                              ON GENERIC LETTER 96-06 i
l                                                                                                                              presented by:
l                                                      NEl GL 96-06 ISSUE TASK FORCE
!                                                                                                                              Dave Modeen, NEl Raub Randels, COMED I
John Minichiello, COMED j                                                                                                                Steve Greco, WEPCo i                                                                                                                            Avtar Singn, EPRI l
!                                                                                                                                April 30,1997
  ,                                                                                                                              Rockville, MD j                                                                                                                                                                                    ATTACFJ1ENT 4 i
 
MEETING OBJECTIVES Discuss status of licensee responses to GL 96-06 and NRC staff concerns and resolution p'an
  - Generic Communication?
  - Formal letters to licensees? When?
  - Other?
Obtain clarification on NRC staff views regarding application of ASME Section Ill, B31.1 and B31.7 requirements and guidance in resolving nonconformances identifiec during licensee review of Generic Letter 96-06 issues.
2 hn
 
4 MEETING OBJECTIVES (contd.)        .
Obtain NRC staff views on the scope of, and schedu e of, any licensee-specific i                and EPRI-sponsored collaborative
;                experimental and analytical efforts to assess:
                  - Piping and valve integrity during overpressurization caused by LOCA
                  - Conservatisms in bounding waterhammer analyses and their impact on fan cooler system 4
operability evaluations For example, if a plant performs a waterhammer Loss of Offsite Power test and if resul<:s show that mods are required, can these mods be postponec                :
to the following outage, proviced that system meets operability?                            '
s h
 
I
                                                                                  )
.                                QUESTION r :E t
l A number of NRC staff communications on near and lo'ng term use of ASME Section lll Appendix F, " Analysis Methodology & Limits," have been provided to industry, but some confusion remains with regards to its use. The NRC staff is requested to provide clarification about the following:
                  - If Appendix F is within a plant's licensing basis, is it acceptable for use as a long-term, over-pressurization solution?
                  - If the answer to (a) is yes, can a plant without Appendix F as a part of its licensing basis restart from a planned or unplanned outage while pursuing a license amendment to make Appendix F a part of the
,                    plant's licensing basis?
                  - If a plant is processing a licensing change to add L
Appendix F to its licensing basis, what typ< . of considerations should be included in the submittal?
                  - If a plant has committed to a schedule for making modifications and/or procedural changes as a long-term over-pressurization solution, is it permitted, prior to making the scheduled modification, to pursue a license amendment to make Appendix F a part of the o an"'s icensina 3 asis?
4 hl
 
l, QUESTIONS Icontd.)                          j QUESTION TWO What information does the NRC need to approve the use of ASME lil, Appendix F as a licensing i
basis for piping systems designed to ANSI B31.17                                    -
;              QUESTION THREE Is the following considered an acceptable l                justification for evaluating the failure mode effect of a stuck open relief valve, occurring post l                accident, located between two closed containment isolations valves?
l The reason a relief valve opens is because the i                pressure has increased between the two i
containment isolation vaives. Consequently, due
,                to the tightness of the subject isolation valves, l                the operiing of the relief valve confirms the containment integrity is provided by the isolation valves and associated piping.
5                          I
 
                                                                  )
QUESTION FOUR Set Pressure (LOCA Oh LYh The design codes used by licensees are silent on " provisions" to be taken for relieving the excess pressure. The 1967 edition of ANSI B31.1 does nat address set pressure requirements. Newer editions do by reference to ASME Section Vlli, paragraphs UG-126 through UG-133. The newer versions of ASME Section Ill (W 1979 and forward) also do not address set pressure / relief capacity for faulted conditions.
Does the NRC staff agree that it is acceptable for a plant to set thermal relief valves at a pressure high enough to prevent unnecessary actuation during normal operation, but low enough to meet current ASME Section Vill (ANSI B31.1) or ASME Section 111 guidance? This option does not propose revising system de sign pressures, but does propose reevaluating the components for the maximum (set plus accumulation) faulted condition pressure.
s hi
 
QUESTION FOUR (contd.'s The NRC staff is also requested to respond to the following specific questions:
l      ANSI B31.1 Applications:
        - Does the NRC staff agree that it is permissible to use a relief valve set pressure equal to 1.09 Pa, where Pa equals the pipe or valve rated pressure, assuming all components sub ected to that pressure are shown acceptable under the faulted condition loading?
ASME Code Applications:
        - Does the NRC staff agree it is permissible to use a relief valve set pressure equal to 0.91 P*, where P*
equals the lower of 2 Pa (Pipe) or 1.5 Pa (Valve),
assuming all components subjected to that pressure are shown acceptable under the faulted condition loading?
        - The basis of the 1.09 and 0.91 factors is a 10%
accumulation, bringing the maximum pres.sure to 1.09 x 1.1 = 1.2 Pa (permitted by B31.1 and Section Vill) and 1.1 (.91) = P* (permitted by Section Ill).
7 hl
 
QUESTIONS              .
l QUESTION FIVE 1
l If waterhammer and/or two-phase flow were not                  l specifically addressed during initial plant / system            l design and licensing, can an analytical approach                I be used without introducing an unrev.iewed safety              l question (USQ) per 10CFR50.59, provided that                    '
the analysis shows that applicable code                        ;
allowables are met once the necessary piping                    l support modifications have been installed?                ,
QUESTION SIX If tests confirm that piping and valves exposed to              l an external temperature increase of up to 250 F are not challenged structurally, will the NRC staff consider accepting the test results as a basis for endorsing an ASME Code Case (relief protection is not necessary when the temperature increase is              !
less than or equal to 250 F) being developed?
l 1
0
 
o C
* EPRl/NPG -
EPRl/ industry Collaborative Project to Support Resolution of GL 96-06 Technical issues Avtar Singh S. Gosselin Mati Merilo H.T. Tang NRCINEllindustry Meeting on GL 96-Oil Rockville, MD l
April 30,1997 iggeve&9 4999' t SARA
;                                    Seekground I
EPRI project (1994-96) on waterhammer prevention, mitigation, and accommodation
                - EPRI reports: TR 106438 and NP 6766 EPRI analysis efforts (early 1996) to predict fan cooler system boiloff and voiding Preliminary plan of the EPRl/ industry collaborative project presented to NRC on December 19,1996 Two collaborative projects to address piping overpressurization and waterhammer/two phase flow issues
                - Two industry meetings (2/25 and 4/9,1997) to review scope of testing and analysis ATTACIN DiT 5 '1
 
o .        ;
i
                                                                                                        ),
EPRUNPO Overpressurization of Isolated Piping Sections Under Postulated LOCA Qplectives
* Develop a credible technical approach to resolve the GL 96-06 containment penetration overpressurization issue and minimize unwarranted hardware                                        ,
modifications a
Develop simplified elastic-plastic methods that can be applied by utilities to evaluate these loading conditions without the need for FEA
                                                                              **"A EPRUNPG Overpressurization of isolated Piping Sections Under Postulated LOCA Approach
                      + Conduct industry survey to define piping configurations, materials, etc.
* Conduct tests to simulate overpressurization conditions Demonstrate piping and valve integrity Benchmark analysis methodology against test data Perform analysis to show how ASME Appendix F criteria can be met
                                                                          - s ARA 2
 
  , o                                                                          1 EPRl/NPO Overpressurization ofisolated Piping Sections Under Postulated LOCA Prolect Scope                                              '
              +
Two-phase project Phase i focuses on scoping (simple'p;pe) tests and analysis to show piping and valve integrity      ,
evaluation meeting Appendix F criteria
              +
Phase 2 provides more quantitative (typical in.
plant configuration) tests and evaluations for generic applications by utilities 4
      ..  .m. .                                                  SARA
                ' EPRl/NPo Overpressurization of Isolated Piping Sections Under Postulated LOCA Phase.1 Utility survey on piping configurations and containment temperature profiles
          +
3 samp!.; piping segments of different D/t (i.e. sch 10, 80
            & 160)
          +
Bounding steady state temperature profile Prost,ure, temperature and piping response history measurement
          +
App endix F evaluation based on measured pressure loading 4
'            - 3 pipirg segments
              - 2 typical valves: gate and globe
      ...m..                                                      SARA 3
 
O v
                                                                                                                                          )
EPRl/NPG -                                                              -
Overpressurization of isolated Piping Sections Under Fostulated LOCA Phase 2 Example of Expanded Test Matrix                                        ,
Configuration              Material          T1"            T2"    Coupon Group
* CG1                CS/SS            1            1        1 CG2                CS/SS            1            1        1 CG3                CSrJ3            1            1        1 CG4                CS/SS            f,            1        1 CG6                CS/SS            't            1        1 CG8                CSISS            1              1  l    1
* Specific test configurations assigned to address geometric lasu6s identified I      in the Phase i survey
                                              " T1 & T2 represent boundine steady state and *ransient teperature proflies                    '
SARA I
EPRl/NPG Overpressurization of isolated Piping Sections Under Postulated LOCA Phase 2 ' cont.)
Appendix F evaluation Valve leakage evaluation If leakage.: potential exists, test to demonstrate leakage and ability to reseal M aterial  Ti  T2 Stra'i; h t P ip e w ith Glo b e V !v e    CS or SS    1    1          ,
Straight Pip e with " sate Valve            CS or SS    1    1 4
 
o
* EPRl/NPG -
Overpressurization of isolated Piping Sections Under Postulated LOCA Technical Products                                  '
* Appendix F analysis methodology Actual safety margins by comparison of analysis and test data Analysis and evaluation guidelines
                < Screening criteria for analysis and evaluation of affected in-plant systems
      ==.m,,                                                  SARA l
EPRl/NPG Overpressurization ofisolated Piping Sections Under Postulated LOCA Deliverables Technical report
                  - Testing report induding testing data
                  - Analysis gtidelines
                  - screening enteria
                  - Inetastic modeis
                  - Appendix F calculations
                  - Long-term integrity enteria Quick ink summary reports Input to ASME code committees
_                                                      -SARA j
S
 
i N
EPRl/NPG Overpressurization of isolated Piping
;                                      Sections Under Postulated LOCA 4
l-                          Preliminary Schedule Phase 1: 2 months after Phase 1 start Phase 2: 8 months after Phase 2 start Final technicul report: 12 months after Phase 2 start i
SARA t
~
EPRl/NPG
* Overpressurization of isolated Piping Sections Under Postulated LOCA PrODOsed NRC -Industry Meetinas Review Phase 1 Results (7/97)
Phase 2 Planning Meeting with the Staff Phase 2 Progress Review Meeting with Staff Review Phase 2 Results and Conclusions (7/98) i l
l 6
 
EPRl/NPG Waterhammer Assessment                    -
                            . For some plants, waterhammer is expected to occur in                          -
the containment air cooler cooling water system during a LOCA/ LOOP scenario The severity of waterhammer, in terms of the loads iraposed on the piping system, is not well understood Analytical methods can give overly conservative results 1
8*
EPRl/NPG Potential Mitigative Factors for Waterhammer Loads-Low system pressure results in small driving 4
forces Bubble vapor pressure opposes acceleration of water slug Heatup and possible boiling in the water slugs prior to impact Air coming out of solution
;                                    - cushioning
                                      - sound speed
                                      - pressure wave absorption s-4 i                                                                                                              7
 
                                                                +    0  e EPRl/NPG
                                                          ~
General Project Objectives
          . Relax conservatisms related to developing thermal-hydraulic forcing functions
          . Provide thermal-hydraulic boundary and initial conditions for waterhammer analysis
          . Assess fan cooler operability with two-phase flow and its ability to transition to design flow conditions SARA EPRl/NPG Proposed Experimental Objectives
: 1. Determine the extemal and intemal heat transfer coefficients for the fan cooler coils,
: 2. Determine the rate and extent of voiding l',- the fan cooler unit and headers during the LOCA LOOP transient.
: 3. Determine the thermal stratification and mixing in the headers.
: 4. Determine the void / liquid interface temperatures and/or the void vapor pressures.
SARA 8
 
                - EPRl/NPG Proposed ExperimeMal Objectives (cont'd.)
: 5. Characterize the effects of the pump starNp transients.          '
: 6. Characterize the effects of the parallel chsanels.
: 7. Characterize the effects of different headet geometries.
: 8. Assess the evolution of noncondenaible Oases and their effect on waterhammer pressure spikes.
: 9. Determine the effects of system components such as throttling valves, surge tanks, loop seals, check valves e.tc.
l
            .- EPRl/NPG                                                      -
Proposed Analysis Objectives Provide a method to apply the results of a testing program to prototypic conditions
                      - e.g. recommendations for heat transfer and fluid flow correlations, as well as system nodalization
                . Support development of experimental facility design, st,aling, test definition, and prioritiration of the test matrix through transient analysis and sensitivity studies.
        . _      .,,                                                    SARA l
9 1
 
EPRl/NPG
                                                                                                          ).
Conceptual Test Facility OPEN LOOP                          CLOSED LOOP          .
J'4          't                                  iti Jh Il
                                                  =J                1.[                .y s.
f                                    y F                9                b                i T              p;--...
_d. Cj
                                                                    % =^h 6
                                                                                  -)
i EPRl/NPG                              -
Deliverables
                                  . Validated thennal-hydraulic models/ correlations applicable to Containment Fan Cooler calculations
                                  . Re.:ommendations/ guidance on how to set up fan cooler system calculations for thermal-hydraulic codes Experimental data
                                                                                    .~
10
 
  ,  e i
EPRl/NPG f                                                                  T Status                    -
                + Participation interest survey:                                .
                      - Approximately twelve to fifteen utilities for the overpressurization project                                  ,
                      - Approximately four to six utilities for the l                        waterhammer project '
l Contractual agreements for funding authorization sent              ,
l                    to interested utilities
                      - Approval contingent on resc'ts being timely in addressing regulatory issues Anticipated start date - May 1997 SARA 11
 
EPRI-INDUSTRY COLLABORATIVE PROJECT                                                              -
TO SUPPORT RESOLUTION OF GL 96-06                                                            ,
4 OVERPRESSURIZATION OF ISOLATED PIPING UTILITY SUR\TY Piping Configurations Valve Descriptions                    ~
Containment Temperature Profiles If                                                                      II Prototypical                                                    Bounding Containment        -
Configuration Groups                                                  Temperature Profiles PWR Designs Piping & Valves BWR Designs (Steady State & Transient) 3 Straight Pipe Segment
                                >                Tests              2 Steady State Bounding Temperature Profile If Burst?            Yes        > Stop I
No Y
cE LOADS st), T(t), E(t)
If                                                                  V PIPE INTEGRITY                                                VALVE INTEGRITY
  ~
Scoping Appendix F Evaluation                                Scoping Appendix F Evaluation (Sample Two Valve Types) y PHASE 1 REPORT Survey & Test Results Prototypical Configuration Groups Temperature Profiles m              Model vs Test Data r          Pipe & Valve Evaluations        '
Phase 2 Test Matrix & Scope laput to ASME Working                                      Presentation 4  Group on Pip.ng Design                        lf          to NRC
                                      '        4A            s
 
EPRI-INDUSTRY COLLABORATIVE PROJECT TO SUPPORT RESOLUTION OF GL 96-06 OVERPRESSURIZATION OF ISOLATED PIPING l  PHASE 2 V
Comnlete Test Matrix Pipe Configuration Groups Steady State and Transient Temperature Probles If
                                    / PIPE LOADS                !
V
                              /              Pm.Tm.c(o /
p                              V VALVE INTEGRITY                                                  VALVE LEAKAGE I      Appendix F Evaluation PIPE INTEGRITY                      Internals Evaluation Appendix F Evaluation V
V Non-Linear Model
                                                      ,                    LEAKAGE Validation                    PREDICTED (Closed Form Solution)                      ,-
V Apply to Other Piping YES
                                        . Configurations V
Configuration                  Coupled Pipe and Valve Test App F Evaluation (Pipe and Valve) 1I                                        NO Screening Criteria Proposed Code Changes                      Limited Valve Leakage & Resen FINAL REPORT Phase 1 and 2 Results Non Linear Model Validation
                            >            Screening Criteria      4 Pipe Integrity Assessments Valve Integrity Assessments Valve Leakage Assessments l
 
o        .
fp2 trag4>k
                  *"                                          UNITED 'TATES
                  ;              ,              NUCLEAR REGULm OAY COMMISSION g              i                          WASHINGTON, o.C. 20666-0001
                    % ,,      #                                  January 24, 1997 MEMORANDUM TO:          DRS Division Directors                        y FROM:                  Gary M. Holahan, Director Division of Systems Safety and Analysis              .
Office of Nuclear Reactor Regulation
 
==SUBJECT:==
STAFF GUIDAEE ASSOCIATED WITH PLANT STARTUP RELATIVE TO ISSUES ADDRESSED BY GENERIC LETTER (GL) 96-06, "ASSUPANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS" The Regions as well as licensees have questioned to what extent the concerns expressed in GL 96-06 constitute restart issues. The purpose of this memorandum is to provide guidance and assure consistency in the staff's follow-up actions associated with the generic letter.
Some licensees have contended that they do not need to perform a preliminary operability assessment, prior to completing the evaluation of the issues as requested by GL 96-06.        It is the staff's position that since information about a potential problem has been disseminated through the Westinghouse Nuclear Safety Advisory Letter (NSAL), Information Notices 96-45 and 96-49, tnd finally through issuance of Generic Letter 96-06, licensees should look at the ganeric issues and determine, consistent with 10CFR Part 50, Appendix B, Criterion XVI, " Corrective Action," if they are susceptible to tne phenomena.
If licensees determine that they are susceptible, they should folic.v the guidance in GL 91-18 and promptly perform an operability assessment.
                      "Promptly" in terms of an initial cursory operability assessment has been interpreted by the staff to mean as soon as possible but generally within 24 hours. In this context, it is recognized that the operability decisior ay only be an initial evaluation and may not be based on a detailed analysis.
However, consistent with the provisions of GL 96-06, it is the staff's expectation that detailed analyses will be completed and reflected in the-information that licensees have been requested to sub. nit to the NRC oy January 28, 1997. See the attached excerpts from GL 91-18 for additional guidance regarding operability assessments.
Snould a licensee determine that plant systems are susceptible to one or more of the phenomena discussed in GL 96-06, the licensee must determine whether or not the system is operable and take actions as necessary to comply with NRC regulations and license requirements. Licensees are not precluded from
 
==Contact:==
James Tatum, SPLB I
415-T305
* This issue is somewhat related to Region III TIA 96-0383.
(2gg fy=d-- g
 
DRS Division Directors                                                                        i      strrting up from an outage or from continued operation with degraded conditions as-long as the affected systems and components are operable NRC regulations and license conditions are satisfied, and the degraded condition is not in conflict with Technical Specification requiremer.ts. If this is not the case, the licensee should follow their existing requirements or make the appropriate submittals for staff review and approval (exemption request, TS change request, etc.).                                                                              -
Specific guidance for judging operability of piping, pipe supports, support                                S plates, and anchor bolts is discussed in the enclosure to GL 91-13. Upon t
discovery of a nonconformance associated with piping and pipe supports, the NRC staff has allowed licensees to use the criteria stated in Appendix F of Section III of the ASME Code for operability determinations. Should licensees determine that piping and supports ure degraded tut operable using this interim criteria, plant restart from an existing outage aay be appropriate depending on the safety significance of the degraded condition and if additional time is needed for engineering, procurement, planning, scheduling, and implementation of design modifiestions (i.e., if the current outage does not constitute the first practical opportunity to take corrective action).
The staff expects that in most cases, licensees will restore affected piping and supports to the criteria stated in the Final Safety Analysis Report for the facility during the next refueling outage.                                                        -
Questions concerning the above guidance should be directed to James Tatuni of my staff at (301) 415-2805 or at F-mail address JET 1.
 
==Attachment:==
As stated cc:    F. Miraglia A. Thadani R. Zimmerman B. Sheron
 
                                                                                                      ._ l l
l Attachment With respect to            t,,^  .bility assessments, Generic Letter 91-18 states:
Section
 
==4.0 BACKGROUND==
1
                "...Without any information to the contrary, once a component or system is established as operable, it is reasonable to assume that the component or
              -system should continue to remain operable, and the previously stated, verifications should provide that assurance. However, whenever the ability of a system or structure to perform its specified function is called into question, operability must be determined from a detailed examination of the deficiency.
The determination of operability for systems is to be made promptly, with a timeliness that is commensurate with .he potential safety significance of the issue. If the licensee chooses initially not to declare a system inoperable, the license must have a reasonable expectation that the system is operable and that the prompt determination process will support that expectation.
Otherwise, the licensee should imediately declare the system or structure inoperabl e. . . "
Set. tion 6.1      Scone and Timino of Ooerability Determinations
                  ...Once a degraded or nonconforming condition of specific SSCs is identified, an operability determination should be made as soon as possible consistent with the safety importance of the SSC affected. In most cases, it is expected that the decision can be made imediately (e.g., loss of motive power, etc.).
In other cases it is expected the decisian can be made within approxima+c1                24 l              hours of discovery cven though complete information may not be available.
!              Some few exceptionkl cases may take longer. For SSCs in TS, the Allowed l            Outage Times (A0Ts) contained in TS generally provide reasonable guidelines for safety significance...
An initial determination regarding operability should be revised, as i            appropriate, as new or additional information becomes available..."
1 Section 6.8          "Indetermirate" State of Ooerability
                  ...Wher 'a licensee has cause to question the operability of an SSC, the operability determination is to be prompt; the timeliness must be commensurate with the potential safety significance of the issue....." Indeterminate" is not l            a recognized state of operability..."
Section 6.13 Pipino and Pine Suonort Recuirements
                ...Upon discovery of a nonconformance with piping and pipe supports, i            licensees may use the criteria in Appendix F of Section III of the ASME code i
'            for operability determinations. These criteria and use of Appentiix F are valid until the next refueling outage when the support (s) are to be restored to the FSAR criteria....Where a piping support is determined to be inoperable, a determination of operability should be performed on the associated piping system."
                              ,                            cr          . .--
 
                    *,                          UNITED STATES j              j            NUCLEAR REGULATORY COMMISSION
      'o            !                      WASHINoTON, D.C. 2068H001
        "% ,      #                                                    December 26, 1996                      %
MEMORANDUM TO:      David B. Matthews, Chief
                                                                                                    #re*U.
a 5
Generic Issues and Environmental                                                4 Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation FROM:                Stewart L. Magruder, Project Manager b y                              6 Generic Issues and Environmental Projects Branch                        /L R Q L Division of Reactor Pregram Management Office of Nuclear Reactor Regulation
 
==SUBJECT:==
 
==SUMMARY==
OF DECEMBER 19, 1996, MEETING WITH THE NUCLEAh ENERGY INSTITUTE (NEI) REGARDING INDUSTRY RESPONSES TO GENERIC LETTER 96-06 On December 19, 1996, representatives of NEI, the Electric Power Research Institute (EPRI), and several utilities met with representatives of the Nuclear Regulatory Cosmiission (NRC) at the NRC's offices in Rockville, Maryland. Attachment 1 provides a list of meeting attendees.
The purpose of the meeting was to discuss (1) the status of industry responses to Generic Letter (GL) 96-06, " Assurance of Equipment Operability and i
Containment Integrity During Design-Basis Accident Conditions," (2) the NRC responses to questions asked at a meeting with NEI on the same subject on October 29,1996, and (3) potential long-term acceptance criteria.
During opening remarks by NRC staff representatives, it was noted that copies of several relevant documents were available for reference by the meeting attendees. These documents included: GL 96-06; Generic Safety Issue (GSI) 150, Rev. 1 (excerpted from NUREG-0933, June 30, 1995); a letter to Mr. Ashok Thadani dated November 8,1996, frem Mr. Roger Reedy; a letter from Mr.
Thadani responding to Mr. Reedy dated December 4, 1996; and a summary of the October 29, 1996, meeting dated November 22, 1996. All of these documents are available to the public.
The NEI representatives began their remarks by stating that licensees are working hard on their responses to the GL. They stated that some licensees are working together and with EPRI as they develop their operability evaluations. The NEI presentation material is included as Attachment 2.
The first topic of discussion was a request by NEI for a clarification of the answer to a question asked by the industry during the Octcber 29, 1096, meeting. The question was related to non-safety systems, which penetrate containment that are not required post-accident, and whether it is acceptable to allcw overstressing and potential pipe yielding in these systems caused by trapped fluid expansion, while declaring containment integrity (Question 11 of October 29, 1996, meeting summary). Tne staff's response to Question 11
: 6. sh vI ~n l (i d I,      M C ''
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D. Matthews                                            Dec d er 20 W stated that overstressing and potential pipe yielding may be acceptable for such systems as long as containment integrity is assured in accordance with the plant design basis. During the meeting, the staff-questioned the reference to non-safety related-systems which penetrate containment and whether they are reclassified in the segments penetrating the containment boundary and between tha isolation ra'ves. NEI representatives affinned that piping in such systems is typically reclassified as safety-related in the segments between the isolation valves.
The next topic of discussion was the applicable acceptance criteria for piping between containment isolation valves-which contains fluid that is subject to thermal expansion. The industry sieted that they believed that the majority of the cases will involve American Society of Mechanical Engineers (ASME) Level D events. The NEI representative asserted that no acceptance criteria exist for this situation and presented an ASME Code Interpretation nl 96-32, dated December 17, 1995, to bolster their argument. The letter from'ASME is included as Attachment 3.-
The NEI representative next stated that, based on the response to the ASME Code Interpretation, the industry believes that a strain-based acceptance criteria should be considered by the staff. NEI and EPRI representatives suggested a strain-based criteria equal-to approximately one half the uniaxial failure strains for carbon and stainless steels. They suggested that piping operability could be maintained with a strain of 10 percent for carbon steel and 25 percent for stainless steel. The staff indicated that it was not prepared to respsnd during the meeting regarding the acceptability of the NEI/EPRI proposal.
* In a teleconference wiWNEI on December 23, 1996, the staff indicated that there is no regulatory precedent or technical basis for approving the proposed
  - strain criteria. The staff noted that ASME Code Interpretation NI 96-32 contains language indicating that Section III of the ASME Code does not provide criteria for pipe deformation resuiting from exposing isolated sections of ASNE Class 1, 2, or 3 piping, containing liquid, which is exposed to an, external heat source that causes thermal expansion of the trapped liquio during a postulated Level D event. The staff notes that the ASME Code contains a pressure stress limit for a postulated Level D event. The ASME Code Interpretation NI 96-32 did not address this pressure stress limit.
On the basis of the ASME Code Interpretation NI 96-32, NEI argued that a definition of the acceptance criteria for evaluating the event is necessary.
NEl requested the staff endorsement of the use of strain criteria to evaluate the piping for the thermal expansion of the trapped fluid. As stated above, the staff believes that the ASME Code does contain criteria that is applicable to the stress produced by internal pressure due to a l_evel D event. The staff further believes that most licensees have specific load-combination criteria specified in the facility FSAR that is applicable to the Loss-of-Coolant Accident (LOCA) or Main Steam Line Break (MSLB) scenarios.      Consequently, the staff is not prepared to endorse a generic strain limit for the GL 96-06 evaluation.
 
D. Matthews                                                -3 December 26, 1996 The staff notes that nonmandatory Appendix F to Section III of the ASME Code contains criteria for inelastic analysis of components'for Level D events.
Depending on the licensee's load criteria _and the facil_ity code of record, the provisions of Appendix F may be appropriate; however, each licensee may need to assess their situation to determine if a license amendment would be necessary.
During the December 19, 1996, meeting, in response to a question from the staff, an industry representative stated that for the vast majority of cases, there are no thermal reliefs in the piping between containment isolation valves. It was also stated that, regardless of whether thermal expansion of fluid between containment isolatior, valves was considered in the design of a plant, the industry agrees that it is a safety issut which must be adequately dispositioned.
A short discussion followed regarding whether the thermal stress should be classified as a primary or secondary stress. An industry representative stated that they have been using the bulk containment pr. essure-temperature curves following a loss of coolant accident as the input to the piping stress calculations. The industry argued that since the conts'nment temperature peaks within a few hours of the pc 11ated accident and then ramps down, the stress should be cor.sidered self-limiting and therefore a secondary stress.
The staff stated that this was not the correct interpretation of the tem self-limiting. - After further discussion, it was agreed that the stress is a primary stress.
The next topic of discussion was GSI-150, "Overpressurization of Containment Penetrations." The NEI representative asked the staff to explain the relevance of GSI-150 to the-issues raised in GL 96-06. The NEI representative noted that GSI-150 had been pieced in the " DROP" category in 1991 by the NRC Office of Research, leading some in the industry to believe that the issue had been resolved. The NRC representatives explained that GSIs are prioritized based on probabilistic risk insights and staff resources. The staff' stated that GSIs placed in the DROP category are sometimes reopened based on new information. The NRC staff is considering reopening GSI-150. There was general agreement that the short term resolution of the issues in GL 96-06 should not be tied to GSI-150.--
The EPRI representative next presented a summary of current EPRI efforts to support resolution of the containment fan cooler waterhammer issue and proposed EPRI efforts related to overpressurization of isolated piping sections. The presentation materials are provided as Attachments 4 and 5.
The EPRI representative stated that EPRI is attempting to prove ti.e hypothesis that the effect of waterhammer may be mitigated in this situation since the piping is subject to continuous heating and the cold water enteri .1 the pipe will be heated to some extent before it contacts the steam voids. The EPRI representative stated that the efforts are just getting underway aad that it appears that the waterhanner analysis will be very plant-specific. With regard to the piping overpressurization issue, the EPRI representative stated
        ,that the presentation material provides very preliminary thoughts and that any follow-up work must be approved by the industry. The NRC staff stated that a
 
December 16, 1996 D. Matthews                                            -
the EPRI efforts were interesting but cautioned the industry not to rely on them until they have been fully developed and reviewed.
The meeting closed with a short discussion of the indu'stry responses to              \
GL 96-06. The industry representatives stated that they expected that all of            X the responses would be submitted by the 120-day deadline and that smet of the responses would probably contain caseitments for additional work. The staff er.couraged the industry to coordinate with each other as much as possible.
The staff also noted that the responses to the GL should include evalcations          i of non-safety related systems as well as safety related systems.
Project No. 689 Attachments: As stated cc: See next page l
                                                                                          \
f
 
NRC/NEI MEETING ON GENERIC LETTER 96-06 LIST OF ATTDDEES December 19, 1996
                                                                                  ~
N821E                                                  ORGANIZATION David Murphy                                            Bechtel                                    '
Bob Hammersley                                          FAI Roger Hayes                                              Southern Nuclear William Birely                                          New York Power Authority Thomas Wroblewski                                        Wisconsin Electric John Anciaur                                            Wisconsin Electric James Hallenoeck                                        PECO Energy                                N Bill Peebles                                            Sargent 1 Lundy Skip Denny                                              PECO Nuclear John McCann                                              Maine Yankee Jorge del Mazo                                          Pacific Gas & Electric Subhash Khurana                                          Florida Powet & Light Glenn Adams                                              Wisconsin Electric    '
Steven Greco                                            Wisconsin Electric Eric May                                                Virginia Power David Stellfox                                          McGraw-Hill Nancy Chapman                                            Bechtel/SERCH Barry Sullivan                                          NUS Dave Modeen                                              NEI Kurt Cozens                                              NEI Avtar Singh EPRI William LeFave                                          NRC/NRR James Tatum                                              NRC/NRR Cheng-Ih Wu                                              NRC/NRR John Fair                                                NRC/NRR Gary Hamer                                              NRC/NRR Beth Wetzel                                              NRC/NRR Kamal Manoly                                            NRC/NRR Dick Wessman                                            NRC/NRR Tad Marsh                                                NRC/NRR George Hubbard                                          NRC/NRR Bill Long                                                NRC/NRR Stu Magruder                                            NRC/NRR Michael Markley                                          NRC/ACRS Attachment 1 l
1 l
 
                                                                                        ~
GL 96-06 Presentec by Kurt Cozens NEl December 19,1996 1
I Attachment 2
 
OBJECTIVES Jndustry status Obtain c!ariication of NRC resconse to question askec at the Dallas meeting              .
Define long-term acceptance criteria 2
di
 
i                                                                  l t
INDUSTRY STATUS e,e Licensees are c'eveloping operability eva uations Typical 120 c ay responses
        - Describe basis for operability    .
        - Discuss actions that will extend past 120 days Evaluation 'of acceptance criteria (code and regulatory)
Evaluation of need to make modifications EP RI efforts to support GL res aonses l
 
e CLARIFICATION REQUEST Question 1' (page 6) states, "Over stressing anc' potential yielding may be acceptable 'or this system as long as containment integrity is asSurec ..."                          -
          - Does this mean a system's containment penetration between the two isolation valves
!            cat he subjected to an over stress and l
potential yielding caused by trapped fluid expansion as long as containment integrity and p st accident functionality are maintained?
Marginsi strain above yield does not impact interrity of the piping system.
4 ha
 
ASME CODE CRITERIA NC-3612.4(a)(3)(e)            " Adequate-consiceration shall be given to the control of fluid pressure caused by heating of the trapped fluid between two valves."
ASME Code Interpretation NI 96-32
    - Question: Does the ASME Section 111, Lision 1, Code provide criteria for evaluating pipe deformation resulting from
,    exposing isolated sections of ASME class 1,
.      2, and 3 piping, containing liquid, which is
,      exposed to an extemal heat source, that causes thermal expansion of trapped liquid during a postulated Level D event.
    - Response: No.
 
ASME CODE CRITERIA (Continuedf .
Jefinition of acceatance criteria is neecec
    - A strain based criteria is appropriate
    - Precedence exist
                                                ~
ASME Section 111, Appendix F GSI-150 Definition of ASME Level D event assurned distortion occurs and replacement is likely SRP, NUREG 0927 (functionality)
 
ANSI B31.1 CRITERIA Acceptance criteria for accident conditions are not clear                                                                                                                                    .
Design pressure is defined as
      " maximum sustained operating                                                                                                                              ~
pressure" Definition of acceptance criteria is needed
            - A strain based criteria is appropriate t
:t    ___-_                - - - - - _ - - _ - - - - - - _ - - - - - - - - _ - - - - - - - - - _ - - - - - - - - - - - - - - - - - _ - - - - - - - - _ - - - -
 
GENERIC SAFETY ISSUE RELEVANT TO GL 96-06 GSJ - 150, "Over Pressurization of Containment Penetrations"                            '
          - Resolved 1995 Priority - DROPPED
          - NRC staff{{letter dated|date=December 4, 1996|text=letter dated December 4,1996}} a Local strains near penetrations may be higher NRC staff reassessing GSI-150 technical evaluation
        - NRC staff requested to discuss a Relationship to GL 96-06 GSI-150 action plan Status of GSI-150 reassessment Basis to use GSI -150 for resolution of GL
 
  .                                                        1 1
CONCLUSION
.              A strain basec' acceotance criteria is appropriate for G _96-06 Long-term resolution                .
l l
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the= =d sensam                                                                    n a mens m e m ar si i            Meehemmet togemens
                                                          .a:". . ,st. .-
awam has1We IrfitEPT 1
December 17,199b                                                '
Mr. Jack I'. Cole Mad Deep PE24 P.O. Bow 968 Mkhlant WA 99352 tiutsort                      Sectaen TA, Divmon L NC/hD:471.2:mid Expansion Effects 0996 Editiosc L'-.- 4                        Yourletter dated Dessaber 3,1996 File:                        NP 5 32                                                                      '
Our understanding of the question and our reply an as follows:
Queston:          Does the ASME section E Dinmon L Code pewide criteria for waluating pipe defonnacon resulting froen esposmg platssi ~% of ASME class 1,2 or 3 prping, tentainsig ikpad, which is expoemd to an ewassmal      heat source that causes thermal expanoson of the trapped liquid during a postulated Imet D ever2?
Reply:            No.
bspec. fully.
                                      ^^
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Nathanlane Secuan EITechnicalInquiay Commitane Q12) 705 7005 27 *** '' -              7. *L~".**.".""' * ~~C' ,,";'?";;~* ;".:~,.".*L'"X ". '".*J** w T '~ ~. M
                                                                                                              'OTR P.02 i
Attachment 3 s
 
l i
l EPRLNPo Proposed EPRl/ Industry Collaborative Effods to Support Resolution of GL 96-06 Technicalissues Avtar Singh Nuclear Power Group EPRI NRC/NEl/ industry Meeting on GL 96-06      -
Rockville, MD                          .
(
December 19,1996 EPRl/NPG Topics
* Current EPRI efforts to support resolution of the Containment Fan Cooler system waterhammer issue
                    + Proposed plan to support resolution of the GL 96-06 long term technicalissues
* ARA Attachment 4
 
  -,    e Current EPRI Efforts            .
* Thermal hydraulic analysJs of CFCU bolloff during LOCA / LOOP events (GOTHIC and RETRAN).
                                                  - timing and extent of volc'ing
                                                  - thermal stratification in the headers
          ,                                      - reduced potential for waterhammer a
EPRI project on Waterhammer Prevention, Mitigation and Accommodation                                      ,
                                                  - Potentialwaterhammer meche.nisms
                                                  - EPRI reports: TR 106438 and NP 6766 SARA 5
EPRutJPo Generic versus Plant Specific Aspects
                                        + Preliminary in.7tmation ir.dicates vast variability in design and operational characteristics
                                                    !ssue resolution is expected to be mostly plant specific
                                        . However, certain CFCU performance related aspects may be genede
                                                  - Two phase flow and voiding
                                                  - Potential waterhammer phenomena S'""
l
 
EPRLWPG                                                .
Generic Technical Tasks to Support Resolution
: 1. Gather input data on plant specific design, configuration, and operational characteristics
: 2. Develop guidelines for realistic assessment of boiling potential and vo! ding for waterhammer evaluation
  ~                3. Perform scaled experiments to generate test data on expected two phase phenomena
: 4. Validatic., of analysis tools via experimental data SARA EPRINPG Summary and Conciusions Generic tools exist to analyze and assess the impact of issues identified in GL 96-06
                      + Containment fan cooler systems have plant specific designs                                                              !
* Genetic phenomena investigations can help                                      1 plant specific resolution Effective resolution can be cchieved via industry collaborative efforts SARA.
                - .m .
                                  -                  .    .      .      -        . . _ .  .= __
 
6    0 e
EPRI NPG Overpressurization ofisol5rted Piping Sections H.T. Tang Avtar Sin 0h Electric Power Research institute
          ~
NRC/NEl/ Industry !!seeting on GL 96-06 Washington D.C.
December 19,1996 g - - ..
EPRI NPG Overpressurization of Isolated Piping Sections Physic ji Behavior
                      . Isolated piping under internal pressure is st essed both in the hoop and axlal direction
                      . Overpressurization may stress the pipe into plastic state The pipe will expand in the hoop direction along the pipe a short distance away from the two isolated ends Depending on the state of strain and geometric nonuniformity (e.g., boundary conditions, th!ckness variations, etc.), the pipe might balloon at a particular_ section
                    ,n w w .a t
Attachment 5
 
EPRINPG Overpressurization of Isolated Piping Sections Physical Behavior (continued)                                            -
                                                                                + Finite plastic eFMnsion of pipe Willlead to precsure reduct on The reduced pressure interaction will reach a state of convergence whereby the pipe will likely be stressed without further plastic deformation Nuclear piping is ductile and thus its ultimate behavior is controlled by strain p,== - . .
EPMINPG
                                                                        ! Overpressurization of Isolated Piping Sections
                                                                          . Strain Criteria Piping performance under overpressurization should be evaluated using strain criteria Uniaxial strain limit of carbon steel is > 20% and stainless steel > 50%
                                                                              - Depending on the state of stress bisaislityttriasiality,the unlauset strain limR rney be reduced by 40 to 50%
                                                                              - Piping operebseity should be maintained wth a strain of >10% for carbon steel and > 25% for stelniess steel Depending on the state of strain and geometric nonuniformity (e.g., boundary conditions, thickness variations, etc.), the pipe might balloon at a particular
                                                                                                                    ~
section GSI 150 uses strain criteria for evaluation r__ ._ . .
 
O                  6
.                          NEI                                                                                            Project No. 689 cc:  Mr. Palph Beedle                                    Mr. Thomas Tipton, Vice President Senior Vice President                                Operations          ,
and Chief Nuclear Officer                        Nuclear Energy Institute Nuclear Energy Institute                            Suite 400 Suite 400                                            1776 I Street, W 1776 I St eet, W                                    Washinton, DC 20006-3708 Washington, DC 20006 3708                                                                          '
Mr. Alex Marion, Director                            Mr. Jim Davis, Director Programs                                            Operations Nuc  ear Energy Institute                            Nuclear Energy Institute Suite 400                                            Suite 400 1776 I Street, W                                    1776 I Street, W Washington, DC 20006-3708                            Washington, DC 20006-3708 Mr. David Modeen, Director                            Ns. Lynnette Hendricks, Director Engineering                                          Plant Support .
Nuclear Energy Institute                              Nuclear Energy Institute Suite 400                                            Suite 400 1776 I Street, W                                      1776 I Street, NW Washington, DC 20006-3708                            Washington, DC 20006-3708 Mr. Anthony Pietrangelo, Director Licensing Nuclear Energy Institute Suite 400 1776 I Street, W Washington, DC 20006-3708 Mr. Ronald Simard, Director Advanced Technology Nuclear Energy institute Suite 400 1776 I Street, W 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
__          .          _                                      . ~ . _ . . _ _ . .                        ,.          __. __
 
o  -
y f ... ,\                              UNITED STATES
                ]          NUCLEAR RE ULATERY GOMMISSION WASHINGTON, D.C. 300064eM
    %* . . .
* lg                        November 22, 1996
                                                                          ~
MEETING SPONSOR:    NUCLEAR ENER'iY INSTITUTE (NEI)
 
==SUBJECT:==
MEETING WITH NEI AND LICENSEES TO DISCUSS GENERIC LETTER (GL) 96-06, " ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT          -
CONDITION 5" On October 2g, 1936, members of the NRC staff attended an industry meeting hosted by Nf! to discuss issues pertaining to GL 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," in Dallas, Texas. The NRC staff participated in the morning portion of the meeting, which was publicly noticed. The afternoon portion of the meeting was open to industry participants only. Attachment I contains a list of questions and answers that the NRC addressed during the meeting.
Attachment 2 is the meeting handout, which includes overhead slides used during both the morning and afternoon sessions and a list of meeting participants.
Representatives from the NRC made presentations concerning the generic letter, including an overview of the generic letter and NRC expectations for licensee responses to the generic letter. The overhead slides used during these presentations are included in Attachment 2.            Following the prepared presentations, the NRC staff addressed questions from the meeting participants regarding GL 96-06. Attachment I contains questions that were received by NRC before the meeting for discussion during the meeting and questions tht.t were presented to the NRC staff during the meeting and the answers to both. The attached clarifications ao rot change the scope or requested actions of GL 96-06.
The staff appreciates the opportunity to clarify and respond to industry concerns regarding GL 96-06. If you have any further questions, please contact Beth Wetzel, the lead Project Manager for GL 96-06, at M1 A15-1355.
L.,    4-  $'      L Ledyar      B. Parsh, Chi (f Plant Systems Branch Office of Nuclear Reactor Regulation Attachments: As stated (2)
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l QUESTIONS AMD ANSWERS FROM OCTOBER 29. 1996 MEETING QUESTIONS GIVEN TO NRC BEFORE THE MEETlHG:                                                                                                            !
Q1.            What are the implications with respect to the licensing basis or design basis of having a 2-phase flow in the system? There is nothing in the FSAR (Final Safety Analysis Report) that specifies 2-phase flow or singis phase flow.
A1.            Regarding the licensing basis code criteria found in the FSAR, 2-phase flow, which occurs as a result of a licensing basis event, could cause significant dynamic loads that have not been analyzed. Therefore, because the licensing basis codes do require consideration of these loads, they should be included in the GL 96-06 evaluations. Also, two-phase flow conditions can result in the accumulation of steam in other areas                                                        of the thct can  lead cooling water system (safety-related and non-safety-related)d and to complicated waterhammer scenarios which must be considere evaluated.
Aside from dynamic effects and waterhammer considerations, there are also fluid flow and heat transfer effects that must be evaluated. Two-phase flow can result in substantial flow oscillations that can upset previously established system flow balances and the introduction of steam in the                                                                t fluid stream can significantly affect heat transfer capabilities.
Q2.              Is this a start-up issue?
A2.            This question would have to be answered on a plant-by-plant basis. If a licensee were nnt able to make a determination that the affected systems were operable prior to start-up, _then the NRC would expect that the licensee would follow appropriate-techni:a1 specifications (TS) and/or license requirements, which could make this a start-up issue.
Q3.            BWR's-do not rely on containment air cooling post accident. We do not think this applies to BWR's. What kind of information does the NRC need l                                back-to conclude that it doesn't apply?                                                                                            ,
A3.            Even though BWRs [ boiling-water reactors) may not rely on containment air cooling, an evaluation should be made to assure the integrity of the system and containment integrity for the design-basis accident loading conditions.
Q4.            What type of fouling factors should be assumed?
A4.              Absent actual test data for the specific plant cooling units, conservatively low values of fouling should be assumed based on new,                                                              ,.
clean, heat transfer surfaces.
QS.
The GL uses the term " delayed sequencing of equipment," what is meant by this ters?
Attachment 1
 
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,                                                                                                                                              AS. The term " delayed sequencing of equipment" was used in the GL to inform
                                                      -licer. sees that they should not limit their evaluations to design-basis accident (e.g., LOCAs (loss-of-coolant-accidents) or MSL8s (main-steamline The NRC believes-loss of offsite power) fan cooll breaks))
that the potential  concurrent  with a LOOP for waterhammer  in CR ((containment systems may exist due to plant-specific time sequencing of safety souipment during or following accidents and not just due to accidents i
concurrent with a LOOP. For example, if the CFCs continue to transfer heat to the CFC cooling systems following an accident (without a LOOP) and flow through the cooling systems has not been established due to secluencing delays in starting the cooling system pumps, voiding may occur anc waterhammer may occur when actual flow through the system is established.                                                                                    '
With regard to the overpressure piping issue, if a plant installs a new 06.
relief valve, do they have to account for a new potential leak path 7 Can we weigh this against the probability of having an overpressure situation?
A6. Even though a new potential leakpath may be introduced by installing a new relief valve. it may be necessary to int. tall the reliefs. - A design-basis accident conM tion which then causes a thermal overpressurization should be considered ir the design, even if it is of low probability. Otherwise, the thermal overpressurization itself could result in a new leak path and result in loss of system function.
: 07. Can we use PRA and risk-based arguments to say that the system is still-operable when we are doing our operability determinations?                                      '
A7. No, PRA [ probability risk assessment) and risk-based arguments are not allowed 1n determining operability.
Q8. The NRC has referenced NUREG/CR-5520 " Diagnosis of Condensation-Induced Waterhammer," in previous discussions with utilities on the containment fan cooler (CFC) waterhammer evaluations. The NUREG states (in section
                                                    -5.1.6, pg. 62) that calculated waterhammer loads easily estimate upper bounds, but actual loads are usually lower by a factor from 2 to 10.
Does the NRC agree with this statement and would they allow licensees to take some credit for tl.is " excessive conservatisn't" A8. As stated in GL 96+06, tl4 ataff believes that the guidance in NUREG/CR-5220 may be usefd in determining various types of waterhanner loads. However, this document-does not contain official regulatory positions or requirements such as found in regulatory guides or in the 10 CFR regulations. In addition, there are numerous other technical references that may also provide useful information. Further, the staff believes that the NUREG/CR-5220 guidance may be. appropriate for some
                                                    . system configurations and not for others. .It is very difficult to generalize that calculated loads may be excessive. The intent of the GL is that licensees carefully ' evaluate the circumstances that could lead to waternammer and determine the associated loads on a case-by-case basis.
____.a_                                          _ _ _ _ _ _ _ .        _    _ _ ._        ,  _ _ _ . _ _ _ _                ~ .
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: 09. Discussion:
The NUREG goes on to give several examples of why the calculated loads are so much higher than actual loads. The examples are a) cushioning by uncondensed steam or non-condensible gas; b) compliance of the piping, i                      hangers, and mounts; c) oblique impact; d) friction on the water slug; and l                      e) reduction in slug length due to steam breakthrough. One way to address I                      these various reductions in waterhammer loads is to use the Joukowski equation with a sonic velocity of approximately half of the sonic velocity in water with no air or other non-condensibles. The resulting waterhamer loads would then be more accurate relative to actual loads.
Q9a. Does the NRC. agree with using reduced sonic velocities as a means of obtaining more accurate results than can be obtained from the standard waterhammer equations?                                                        i A9a. As stated in this question, there may be actual phenomena associated with
,                      any actual event that may reduce the magnitude of actual loads con. oared to  )
calculated values. However, there are substantial uncertainties involved in predicting these conditions, making it difficult to ensure that favorable conditions will exist to ensure such reddctions. Typically, good engineering practice involves making assumptions that provide            i reasonable assurance of a conservative assessment.                            !
Q9b. Has the NRC published any additional information on methods to more accurately calculate waterhammer loads since NUREG/CR-5520 (dated 10/88),
especially low-temperature, low-pressure waterhammer loads?
:              A9b. The NRC has not issued any other technical guidance since the issuance of NUREG/CR-5220 regarding waterhammer loads.
Q10. Some actual plant-snecific test data is available for CFC waterhammer events. This data confirms the statement in NUREG/CR-5520 that calculated loads are 2 to 10 times higher that actual loads.
Would the NRC accept actual plant-specific test data as an alternate means of determining waterhammer loads and resulting impact on plant systems and structures, in lieu of calculated loads and impacts?
A10. The NRC would accept actual plant-specific test data as a means of determining actual waterhammer loads if the test data are representative of the operating or accident condition being evaluated.
l Q11. In the tiL's backfit discussion for the overpressurization of piping, there are statements regarding meeting the ASME code. ASME experts have said that there is nothing in the ASME code precluding this condition. Fluid systems are allowed to temporarily overpressure and that the comment about code applicability is incorrect here.
Can the NRC coment on this?
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4 All. The staff does not agree with the assertion that the codes do not require any consideration of this :,verpressure. The ASME Code has different stresa allowables depending on the service loading conditions. Licensees are expected to use the appropriate load combinations and stress limits.
QUESTIONS ADCRESSED TO NRC DURING MEETING:
Ql.              Is there an implementation date that NRC expects this to be completed by?
A1.              No, the NRC h a not specified an implementation date by which licensees should complete any corrective actions. Licensees should follow GL 91 18
[*Information to Licensees Regarding Two NRC Inspection Manual Sections on Degraded and Nonconforming Conditions and on Operability") and 10 CFR Part 50, Appendix B, Criterion 16, for evaluation of the operability and implementation of corrective tetions. Timeliness of the fix should be      '
connensurate with the safety significance of the problem.
Q2.              If in the process of developing the 120-day response to the GL the licensee determines that 120 days will not be sufficient time to complete its response.
* hat is the appropriate action?    ,
A2.
Any possible problems with meeting the 120-day response shoeld be promptly communicated with appropriate basis to the NRC project manager for that plant, who will discuss the particular scheduling problems with the technical reviewers for GL 96-06.
Q3.              (paraphrased from several questions on the same subject)
If the plant is licansed for a LOOP coincident with a LOCA only, is it mandatory to consider a LOOP after a LOCA?
A3.
No, it is not mandatory to consider a LOOP after a LOCA if the plant is licensed for a LOOP coincident with a LOCA only. A licensee is not required to go beyond its licensing basis.
Q4.              Is this an " operability" concern or an ASME qualification issue?
A4.
The issues discussed in GL 96-06 are concerns both for assuring operability of systems, structures, and components as outlined in the GL 91-18 guidance and for demonstrating compliance with applicable codes and standards or other licensing-basis requirements.
QS.              Is there a need to consider fan mstdown for infrequent normal operation of fan coolers, such as surveillance?
A5.              No, there is no need w consider coastdown from infrequent normal operation of fan coolers, such as surveillance. Fan coastdown would need to be considered if the fan coolers were used to maintain containment traperatures during normal operation to account for infrequent extreme weather conditions, such as high sunner temperatures, i s.            Can valve leakage be credited as a basis for long term operability?
 
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                                                                              -                              i l        A6.      Considering that this question relates to pressure increases in isolated piping, valve leakage would only be allowed if it is quantifiable, predictable, and known. However, the leakage should be evaluated in terms                    l of impact on other plant equipment performance.
The NRC states totay, " Don't go beyond your licensing basis," however, the                  l
: 07.                                                                                                  '
NRC states that the licensee should consider the effects of returning DW (drywell) cooling to service even though DW cooling is not taken credit for in the licensee's design / licensing basis. This sounds like a contradiction. Please expand on this stance.                              .
A7. A licensee is not required to go beyond its licensing basis for GL 96-06 purposes. However, the NRC considers that the licensing basis includes maintaining containment integrity. Therefore, if returning drywell cooling per your procedures (even though it is not credited for accident mitigation) crt h the potential to lose containment integrity, an evaluation would be required.
Q8. Do you have to consider the pressurization of non-safety loops inside containment and the effects of this pressurization o,n containment isolation?
l          AB. Yes, they need to be considered to ensure that containment integrity is mainteined. Additionally, consideration needs to be given to whether the patential failure would divert flow from safety equipment and prevent the safety equipment from performing its required safety function.
i
: 09. Since BWR4 DW coolers are not safety related or required post accident, you said we must consider operator procedures which use non-safety related coolers post accident, if we use coolers in procedures post accident and they fail, we isolate RBCCW (reactor building closed-cooling water) and move on. What response does NRC want for this? RBCCW has isolations in l                  Appendix J program.
A9. In the response, the licen'see should discuss the fact that the use of the nonsafety equipment was evaluated and that these systems could be isolated and containment integrity would be maintained.
l            Q10. Please expand on your application of single failure requirement as it relates to penetration overpressure.
A10. It is the staff's understanding that in order to address thermal pressurization of piping penetrations (especially those that are isolated by containment isolation valves during an accident condition), one option i                  being considered is to allow the pressure to exceed the penetration piping i                  design pres:ure such that the penetration will crack and pressure will be relieved either inside or outside containment (depending on the crack location), and then analyze the consequences of these penetration failures. Therefore, the above question has been amplified as follows: is this an acceptable approach for licensees to take in addressing the thermal overpressurization issue, and if it is, how would the single-failure criteria be applied.
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_                      ~ _    . . _ , _ ._                  _ _ . . _ .
 
o o It is the intent of GL 96-06 that the integrity of affected system piping be assured, especially for thk piping containment penetrations.
Therefore, the structural integrity of the piping shoulu be assured for any design-basis overpressure condition.
Qll. For a non-safety system that is not required post accident, that has both interior and exterior isolation valves, is it acceptable to allcw overstressing and potential pipe yield and still declare containment integrity?                                                              .
All. The capabilities of systems in terms of their accident mitigation functions should be considered in a broad sense. If a piping system is associated with a safety-related system, and its failure could impact containment integrity, then it should be designed to withstand system overpressure due to thermal expansion or 2-phase flow or waterhammer loads. However, if a nonsafety-related system's failure would have no impact on a safety-related system performance (such was not the case for Maine Yankee), then it need not be designated to withstand these loads.        l However, licensees should e.sure there are no post-accident situations l            when this system would be called on by the operatort or by automatic features ;o function. Overstressing and potential pipe yielding r..y be acceptable for this system as long as containment integrity is assured in accordance with the plant design-basis,                                        i Q12. Can " Leak Before Break" be used to limit containment temperature after a DBA (design-basis accident) on an interim basis in order to suoport an operability assessment for containment air coolers?
A12. Generally speaking, the elements associated with
* operability evaluations" deal with ways of mitigating the particular event in a " defense in depth" mtnner, rather than relying on "best estimate" or 'more probable" initiating event sequences. For example, arguments involving redundant trains or equipment, available operator actions, and time available to take mitigative actions are generally acceptable for making operability assessments, while arguments involving the initiating event probability are not, alone, sufficient. Therefore, relying on ' leak-before-break" is not, alone, sufficient for an operability assessment. Other defense-in-depth, engineering judgment, and compensatory actions arguments should be considered, in addition to arguments involving the initiating event.
Q13. Will a consistert position be articul:.ted to the Regions from NRR relative to startup issues?
A13. We will communicate the results of this workshop in a variety of ways.
First, we will state our position in an upcoming counterparts meeting with the directors of the Division of Reactor Safety for the regions. Second, we will utilize the meeting summary with attached questions and answers to comunicate our positions. And third, we routinely are in touch with the regicns and will discuss our positions. We routinely receive questions from the regions on ways to deal with operability, and we will use these opportunities to state our positions to arrive at consistency.
 
8                      4 1
i                                                                                                                                                                                                                                    is it appropriate to take Q14. With regard to heat ing ' causing voiding,ient for steam condensing on the
.                                                      credit for a reduced heat transfer coeffic                                                                            i I
exterior of the tubes (tratisferring heat) to steam within the tube vs.
steam (condensing on the exterior of the tubes transferring heat) to water (within the tube)? (rewording in parentheses)                                .
3 A14. The staff recognizes that the uncertainty involved in predicting hsat transfer coefficients may be significant. There should be reasonable                                                  ;
!                                                        assurance that the assumed coefficients are conservative for predicting heat transfer. Further, tne accuracy of the assumed values siould be -                                                i consistent with the level of detail of the analysis. That is, for                                                      l l
bounding analyses that have little refinement of the various analysis j                                                        parametn . more conservative coefficients should be assumed to acceunt for ovmil uncertainty in the results.                      For analyses that use more i
accurate correlations based on test data for the various parameters, less 2
conservative coefficients may be acceptable.
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;                                Q15.                    Is drilling holes in the disks of containment isolation valves (inboard disks, inboard valves) a possible solution?
!                                A15.                    lhe drilling of holes in system components could provide adequate pressure relief for a ther;nal overpressure condition, provided the resulting degradation of the structural integrity and saftty functions of the
:                                                        components are adequately addressed.
;                                  Q16. NUREG 5220 statas that combination of waterhammer and seismic loads should not.be required for pipe stress evaluations. Typically, FSARs include piping load combinations of LSCA and seismic that were designed primarily for reactor coolant loop piping. What is the NRC's positinn on the load combinations for service waterhammer?                                                                                l i                                  A16. The licensees should conform to the plant-specific design-basis load combination requirements for the piping being evaluated. For example, where loads in the containment fan cooler system are the result of a design-basis LOCA (including expected waterhammer loads), these LOCA-                                              l l
induced loads should be combined with other loads (i.e., dead weight, I
;                                                          thermal, and seismic) as required by the plant design basis. The plant-specific design-basis load combinations may consist of FSAR requirements,                                          ;
l ASME Code requirements, or other commitments.
Q17. Are air traps and/or standpipes an acceptable solution for mitigating pressure build-ups (since air is compressible)?                                                                    j A17. Air cavities of various configurations may be-acceptable solutions to the                                                            ;
thermal overpressure issue. However, since air (and other gases) are                                                l 1                                                          soluble in water, it must be assured that the air cavity will actually                                            .
exist when needed.                                                                                                  !
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8-Q18. The Beaver Valley overpressurized penetration was due to heating over a long time, but the GL ulludes to heating over a short time in accident conditions. Is it your intent for us to answer for just accident
                                .            conditions or both?
A18. Any potential thermal overpressure condition should be evaluated, including any normal or accident conditions, t
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4 o UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHitt TON, D.C. 20555 0001 September 30,1996 NRC GENERIC LETTER %06:- ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN. BASIS ACCIDENT CONDITIONS Addratista All holders of operating licenses for nu : lear power reactors, except for those licenses that have bten amended to possession-only status.                                                    '
Purnose The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to (1)    notify addressees about safety-significant issues that could affect containmera integrity and equipment operability during accident conditions, (2)    request that all addressees submit certain information relative to the issues that have been identified and implement actions as appropriate to address these issues, and (3)    require that all addressees submit a written response to the NRC relative to implementation of the requested actions.
P*'' Ats.unii As a result of recent NRC inspection activities, licensee notifications, and event reports, several safety-significant issues have been identified that have generic implications and warrant action by the NRC to assure that these issues have been adequately addressed and resolved. In particular, the following issues are of concern:
(1)' Cooling water systems serving the containment air coolers may be exposed to the hydrodynamic effects of waterhammer during either a loss of-coolant accident (LOCA) or a main steamline break (MSLB). These cooling water systems were not designed to withstand the hydrodynamic effects of waterhammer and corrective actions may be needed to satisfy system design and operability requirements.
(2)    Cooling water systems serving the containmerJ air coolers may experience two-
            . phase flow conditions during postulated LOCA and MSLB scenarios. The heat                !
reewval assurr ptions for design-basis accident scenarios were based on single-phase flow conditions. Corrective actions may be needed to satisfy system design and operability requirements.
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GL 96-06 September 30,1996                                i Page 2 of 10
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(3)    Thermally induced overpressurization of isolated water filled piping sections in                              !
containment could jeopardize the ability of accident mitigating systems to perforrn                          i their safety functions and could also lead to a breach of containment integrity via                          :
bypass leakage. Corrective actions may be needed to satisfy system operability                                I requirements.
The sections that follow contain additional background information about each of these                                i issues.                                                                                                  .
Waterhammer On February 13,1996, the Pacific Gas and Electric Company (PG&E, the licensee for Diablo Canyon Units 1 and 2), detennined that component cooling water, which is circulated through the containment air coolers, could flash to steam in the cooler unit cooling coils-during a design-basis LOCA with a concurrent loss of offsite power (LOOP) or with a delayed sequencing of equipment.- This condition was reported to the NRC in Licensee Event Report (LER) 196-005, dated April 26,1996. -                                                                    ;
The Diablo Canyon units have five containment air coolers in each containment, these                                  >
are typically used during normal plant operation to prevent excessive containment temperatures. The containment air coolers are also automatically initiated engineered safety features that are relied upon to help maintain containment integrity by performing their heat removal function during postulated accident conditions. The air coolers in the                          -
Diablo Canyon units transfer heat from the comainment to the respective unit's                                      .
component cooling water system (a closed loop system).
PG&E reported that, dudng a postulated design-basis LOCA with a concurrent LOOP, the component cooling water pumps and the air cooler fans will temporarily lose power (an expected condition). The component cooling water flow stops almost immediately, while the fans coast down over a period of minutes. The first air cooler fan will restart on slow                          '
speed approximately 22 seconds after the LOOP and the component cooling water pumps l                      Will restart 4 to 8 seconds later. In this scenario, the high-temperature containment atmo-                          )
sphere will be forced across the containment air cooler's cooling coils for up to 30 seconds                          ,
with no forced component cooling water flow through the cociers. PGkE determined that the stagnant component cooling water in the containmer.t air coolers may boil and create a substantial steam volume in the component cooling water system.. As the component cooling water pumps restart, the pumped liquid may rapidly condense this steam volume                                  !
and produce a waterhammer. The hydrodynamic loads introduced by such a waterhammer event could_be substantial, challenging the integrity and function of the containment air coolers and the associated component cooling water system, as well as posing a challenge to containment integrity. As corrective action, PG&E has installed a nitrogen pressurization system on the component cooling water head tank to increase the margin to boiling.
On June 20,1996, Westinghouse Electric Corporation issued Nuclear Safety Advisory                                      ,
i.etter NSAL-96 003, ' Containment Fan Cooler Operation During a Design Basis
                    . Accident," to alert its customers to the potential rafety issue that was                                          v
 
D                C GL 9606 September 30,1996 Page 3 of 10 identified by PG&E (Westinghouse is the reactor vendor for the Diablo Canyon units). In i
NSAL 96-003, Westinghouse recommended that licensees review their containment cooling systems to determino if their safety related containment air coolers are susceptible to waterhammer.
]                                                                                                            '
On July 22,1996, the Connecticut Yankee Atomic Power Company (CYAPC, the licensee
;                        for the Haddam Neck nuclear oower plant) declared all four of the containment air coolers at the Haddam Neck p ant inoperable and initiated a plant shutdown in                            .
accordance with Technical Specification requirements. The containment air coolers at the Haddam Neck plant are the only components that are credited for post-accident containment heat removal, and station service water (an open lcop system)is the cooling                                ,
medium for the containment air coolers. The containment air coolers were declared inoperable after CYAPC completed its review relative to Westinghouse NSAL-%-003. The licensee's analysis predicted hydrodynamic loads in the service water system from waterhammer that exceeded piping and support structural limits.
Or August 12,1996, the staff issued Information Notice (IN) 9645
* Potential Common.                            '
Mode Post Accident Failure of Containment Coolers," to alert addressees to the potential failure mode of the containment air coolers and their associated cooling water systems.
IN 96-45 discussed the information that was reported by PG&E and CYAPC relative to the Diablo Canyon and Haddam Neck plants, respectively, and attached a copy of Westing-house letter NSAL 96-003.                        -
Two-Phase Flow in Safety Related Piping and Components In July 1996, the NRC issued Inspection Report 50-213/96-201, 'Special Inspection of 4
Engineering and Licensing Activities at Haddam Neck-Connecticut Yankee" Among other things, the report identified an issue relative to tw& phase flow in the station service water system. The inspection team reviewed the service water system flow models, calculations, and operational data and found that some steam may be produced in the l                      . service water system as the service water flows through the containment air coolers during design basis accident conditions. However, the licensee's service water system model and calculations only assumed single phase flow conditions (liquid phase only) and did not consider two phase flow conditions (both steam and liquid present). The licensee is currently evaluating the system to determine whether or not corrective actions
,                      are needed.
On July 23,1996, the Wisconsin Electric Power Company submitted information regarding two-phase flow in the service water system at the Point Beach nuclear plant during a des!gn-basis LOCA. The licensee's preliminary evaluations concluded that after the cooling water is heated via heat transfer from the containment air coolers, some steam could be formed at the air cooler outlet throttle valves. This two phase mixture (steam and water) would result in a higher frictional pressure drop in the service water return piping cnd would ultimately affect the service water flow and the heat removal capabill-tier, of the containment att coolers. Steam formation due to low pressure and high tempemture in the service water system could reduce the service water flow rates through the containment r.ir coolers to values below those needed to
    . , ~ , . , . .          _ . - , - . - _ . . _ - , ..--..,_,n ,. ,. . . . , _ . - - , .. , , - , .., _ _ - , ,...._rm,.-,w .--m_..    .-,_
 
CL 9606 September 30,1996 Page 4 of 10 satisfy design basis heat removal requirements. The licensee is completing more detailed analyses to determine if immediate corrective action is warranted.
On August 20,1996, the Public S-rvice Electric and Gas Company (the licensee for Salem 1 and 2) notified the NRC of a condition that is not bounded by the existing design basis for the Salem nuclear power plants (EN 30900). The licensee reported that because the service water isolation valves for the nonsafety related turbine loadt do not start to close until approxmtately 30 seconds into the emergency loading sequence, the service water system may not be able to supply sufficient flow for the containment cooling function during accident conditions. The licensee determined that the initial heat transfer rates through the containment air cooiets could result in additional
* flow restrictions" in the air cooler tubes, further decreasing the flow of service water through the containment air coolers as a result of the higher frictional pressure drop caused by two-phase flow. At the time of the licensee's notification, the Salem units were shut down for refueling.
: Overpressu?izatlon of isolated Piping Sections On July 3,1996, Duquesne Light Company (the licensee for Beaver Valley Units 1 and 2) notified the NRC that during surveillance testing of a component cooling water inlet l valve to the RHR heat exchanger on Unit 1, the motor operated butterfly valve located I inalde the containment would not open (EN 30833). The licensee found that pressure in the piping section between this valve and a closed manual butterfly valve located outside the containment measured slightly higher than the system design pressure. After the pressure in this isolated sectit,n of piping was relieved by opening a drain valve, the remotely operated butterfly valve was opened without any troub!e. The licensee concluded that pressure in the isolated section of piping increcsed when the trapped water was heated up by increased ambient temperatures. The section of piping was isolated in the spring when the unit was shut down and ambient temperatures were much lower than temperatures that existed in the summer after the plant was returned to power operation and ambient temperatures reached about 32 *C [90 'F).
On July 19,1996, the Maine Yankee Atomic Power Company (MYAPC, licensee for the Maine Yankee nuclear plant) notified the NRC of a condition that was outside the plant design basis (EN 30769). De primary component cooling water (PCCW) system at the Maine Yankee plant has a nonsafety-reinted subdivision that serves the containment fan coolers (not needed for accident mitigation), and a safety related subdivision that serves ECCS equipment. The nonsafety-related subdivision of PCCW has a swing-check valve at the containment inlet (supply) penetration, and an air operated valve at the containment outlet (return) penetration. During a design basis LOCA, the contairunent isolation logic initiates closure of the air-operated outlet valve, thereby stopping the flow of water. The licensee has determined that heat from the containment accident environment could cause the PCCW in the containment fan coolers between the inlet check valve and closed air-operated outlet valve to expand, rupturing this portion of the PCCW system. Water from the PCCW system is then able to flow through the supply check valve for the containment fan v
 
a s GL 96-06 September 30,1996 Page 5 of 10 coolers and out the rupture, rendering the PCCW system inoperable and jeopardizing safety-related equipment that is cooled by the safety related division of the PCCW system.
Upon recognizing this postulated scenario, the licensee promptly shut down the Maine Yankee plant. To correct this, the licensee plans to install a press 6re relief valve on each of the six containment fan caoler PCCW branch lines downstream of the supply check valves.
On August 20,1996, the staff issued Irformation Notice (IN) 96-49," Thermally Induce'd Pressurization of Nuclear Power Facility Piping," to alert addressees to the potential for safety.related piping to become overpressurized during accident conditions. IN 96-49 dhcusses the information reported by Duquesne Light Company and MYAPC relative to Beaver Valley Unit 1 and the Maine Yankee plant, respectively.
Discussion The issues discussed in this generic letter pertain to situations that may not be bounded by the applicable system design capabilities and for which corrective actions may be needed to satisfy equipment design and operability requirements. The sections that follow contain additional discussion about each of these issues.
Waterhammer At many plants, containment air coolers satisfy a significant safety function by removing heat from the containment and reducing post-accident containment pressure. The hydrodynamic loads imposed by waterhammer can be substantial, challenging the integrity and function of the containment air coolers and the associated cooling water system, as well as posing a challenge to containment integrity. Waterhammer in cooling water systems associated with nonsafety related containment air coolers can also challenge containment integrity by creating a containment bypass flow path, and interfacing safety-related systems can be affected. During this accident scenario, the ste:.m that is produced in the containment air coolers may accumulate in other parts of the cooling water system, restricting tiow as well as causing waterhammer damage. Plant vulnerability to the postulated waterhammer scenario depends on a number of factors, such as piping configuration, how long it takes for the flow of cooling water to stop, the coastdown rate of the fans in the containment fan coolers, the operating pressure and pressure decay rate of the cooling water system, how long it takes to establish forced cooling water flow, the containment temperature profile, and other site-specific parameters.
The postulated failure scenario is applicable to Mth LOCA and MSLB events that involve a loss of offsite power, a loss of cooling water flow to the containment air coolers (e.g., one train of cooling water inoperable), or the sequencing of equipment that can affect the containment cooling function. Steam formation and waterhammer in cooling water systems associated with cafety-related and nonsafety related containment air coolers may not require a loss of offsite power for this scenario to be valid.
* C GL 96-06 September 30,1996 Page 6 of 10 Two Phase Flow in Safety Related Piping and Components 4
Two phase flow (i.e., both steam and liquid) in cooling water system! associated with the containment air coolers can significantly interfere with the abillt of the containment air coolers to remove heat under design basis accident conditions, and can interfere with the cooling of other safety relateo compontats, nest cooling water systems were designed assuming single-phace flow conditions (i.e., liquid only) and containment heat transfer analyses are based on this assuniption. Two-phase flow is a much more complex situation to deal with analytically than single-phase flow and involves additional hydrodynamic l                                loading considerations as well as flow, heat transfer, systems interaction and erosion considerations. Additionally, the steam that is fonned during two phase flow can
:                                accumulate in the cooling water system, restricting flow and resulting in a waterhammer as discussed above.
l                                Overpressurization of isolated Piping 4                                                                                                                                                                                    1 Because of its thermal expansion, water heated while it is trapped in isolated piping sections is ca sable of producing extremely high pressures, his phenomenon is typically a                                                          ;
design consit erstion. Piping design codes as far back as U.S.A. Standard (USAS) B31.1 (1%7), have explidtly recognized the need to unsider the effects of heating fluid that is                                                          <
,                                trapped in an isolated section of piping. The p9tential for thermally induced expansion                                          # of fluid trapped in valve bonnets was one reason for issuing Generic I etter (GL) 95-07,                                                      .
!                                " Pre.sure locking and Thermal Binding of Safety Related Power Operated Gate Valves."
In addition, several information notices (ins) have been issued discussing the pressuriza-j                                tion of water trapped in valve bonnets, including IN 95-14, " Susceptibility of Contain-ment S unp Recirculation Gate Valves to Pressure Locking," IN 95-18, "Poter.tlal Pressure-
;                                Locking of Safety Related Power-operated Gate Valves," IN 95 30," Susceptibility of LPCI                                                          ,
and Core Spray 1.4ection Valve to Pressure Locking," and IN 96-08, " Thermally induced Pressure locking of a HPCI Gate Valve."
he potential for systems to fall to perform their safety functions as a result of thermally
* induced overpressurization is dependent on many factors. These factors include leak
'~
                                . tightness of valve seats, bonnets, packing glands and flange gaskets; piping and                                                                  '
component material properties, location and geometry; ambient and pcst accident temperature response; pipe fracture mechanisms; heat transfer mechanisms; relief valves and their settings; and system isolation logic and setpoints. Engineering design and modification evaluations, which include systematic evaluation of heat input to systems and components with consideration of factors such as those just noted, can detect                                                                ,
<                                  conditions which may influence system operability under normal operating, transient, and accident conditions.
f' Under the " single-failure concept," failure due to overpressurization does not preclude                                          .
consideration of additional active and passive failures in the same and other systems in evaluating plant response to a postulated acciden' If relief valves are installed to prevent overpressure conditions, consideration must be given to the effects of stuck-open relief i                                valves and associated environmental flooding and radiation hazards,                                                                    v
                                                                                                                                                                                  'I
                                                . , ,                    ,,,..m,. ~ . . ... - . . - .            -..-_, . . . _ .      - . , , - . . , . . - - - ~ . -        --
 
                      .-b        4 GL9606 September 30,1996                                                    ,
Page 7 of 10                                                          ;
                                                                                                                                                                                                          \
Rennented Action (s)-
Addressees are requested to determine:                                                                                                                        't (1)        if containment air cooler cooling water systems are susceptible to either                                  .
waterhammer or two phare flow conditions during postulated accident conditions;
                                                                                                                                                                        ~
(2)- _ if piping systems that penetrate the containment are susceptible to thermal expansion of fluid so that overpressurization of piping could occur.                                                                              ;
in addition to the individual addressee's postulated accident condit!ons, these items should be reviewed with respect to the scenarios referenced in the generic letter.                                                                            :
With regard to waterhammer, addressees may find Volumes 1 and 2 of NUREC/CR-5220, "DMgnosis of Condensation Induced Waterhammer," dated October 1988, informative and useful in evaluating potential waterhammer conditions.
                                        . If hystems are found to be susceptible to the conditions discussed in this generic luter,                                                                      ,
addressees are expected to assess the operabilitypf affe+=d systems and take corrective action as apympdate in accordan'ce'wWtle3quirements stated in)' 4M"PR                                            50 Appendix B and as required by the plant Technical Specifications (GL 91 1                                  Information            _
to Licenrees Regarding Two NRC Inspection Manual Sections on Resol                                      of Degraded                                            .
and Nonconforming Conditions and on Operability," dated November 7,1991, contains                                                                              !
                                      - guidance on the review of licensee operability detenninations and licensee resolution of degraded and nonconforming conditions.
Reauested Information '                                                                                                                                        '
s Within ?20 days of the date of this generic letter, addressees are requested to submit a:                                                                      ;
written summary report stating actions _ taken in response to the requested actions noted                                                                        .
above, conclusions that were reached relative to susceptibility for waterhammer and two-phase flow in the containment air cooler cooling water system and overpressurization of piping that penetrates containment, the basis for continued onerability of affected systems and components .as applicable, and corrective actions that were impfemented or are planned to be implemented. If systems were found to be susceptible to the conditions that are discussed in this generic letter, identify the systems affected and describe the specift:
: circumstances involved.
Eggaired Resoonse Within 30_ days of the date of this generic letter, addressees are required to submit a written '
response indicating:- (1) whether or not the requested actions will be completed, (2)
                                      - whether or not the requested information will be submitted and (3) whether or not the requested information will be submitted within the requested time period. Addressees who choose not to d.,.
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o a GL 96-06 September 30,1996 Page 8 of 10 i
l complete the requested actions, or choose not to submit the requested information, or are              i unable to satisfy the requested completion date, must descdbe in their response any                    !
alternative course of action that is proposed to be taken, including the basis for                    i establishing the acceptability of the proposed alternative course of action and the basis for
,                        continued operability of affected syst#:ms and components as applicable.                              ;
Address the required written reports to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, D.C. 20555-0001, under oath or affirmation, under                ,
the provisions of Section 182a, Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(0. In addition, send a copy to the appropriate regional administrator.
Bgkfit Discussion Title 10 of the Code of Federal Reculations (10 CFR) Part 50 (Appendix A) and plant licensing safety analyses require and/or commit that the addressees design safety related components and systems to offer adequate assurance that those systems can perform their              -
safety functions. Specifically,10 CFR - Part 50,(Appendix A, Criterion 38) specifies a "cystem to remove heat from the reactor containment. The safety function of this system is to rapidly reduce pressure and temperature in the containment following any loss-of-coolant acddent and to maintain them at acceptably low levels." Addition-ally, Criterion 44 of Appendix A spedfies a " system to transfer heat from structures, systems, and components important to safety.- The system safety fr tion shall be to                  .
transfer the combined heat load of these structmes, systems and et mponents under normal operating and acddent conditions." The heat load values as defined in final safety analysis reports are based on single-phase flow assumptions for the containment air cooler cooling water systems. The potential for waterhammer and two phase flow raises concerns that these systems will not meet their design-basis requirements as spedfied in            ;
10 CFR Appendix A, Criteria 38 and 44. Further,10 CFR Part 50 Appendix A, Criteria 1 and 4 specify that safety related systems be designed to offer adequate assurance that those systems can perform their safety functions under acddent conditions. Accordingly, licens-ees are required to ensure that the containment air coolers and their associated cooling water systems that may be affected by waterhammer or by two-phase flow are capable of              '
performing their required safety functions and that containment integrity will be maintained.
Licensees are also required either by their commitment to USAS B31.1 or the American Society of Mechanical Engineers (ASME) Code for piping design or by virtue of 10 CFR 50.55a,~which endorses various editions of the ASME Boiler and Pressure Vessel Code, to comply with design cdterin which spedfy that piping systems which have the potential to experience pressurization due to trapped fluid expansion shall either be designed to withstand the increased pressure or shall have provisions for relieving the excess pres.
sure. The pountial for overpressurization raises concerns that :hese piping systems will
                          - not meet the'r design code criteria.
The actions requested in this generic letter are considered compliance backfits under the provisions of 10 CFR 50.109 and existing NRC procedures to
_ . _ _ _ . . _ . _ _ _                      _    _ _ _ _ ~ . ~ . _ _ _ _            _ _ . . _ _ _ . _ _ _ _ . _ _
 
o''    o GL 96-06 September 30,1996 Page 9 of 10                            >
j                            - ensure that containment integrity will be maintained and that safety related components and piping systems are capable of performing their intended safety functions and satisfying their licensing basis coh criteria, respectively; and that containment integrity
;                            - and these safety related piping systems and components will not be adversely affected by i
the occurrence of waterhammer, two phase flow, or thermal overpressurization that may occur in safety related and nonsafety related systems that penetrate containment. In accor-i dance with the provisions of 10 CFR 50.109 regarding compliance backfits, a full backfit -
i                            analysis was not performed for this proposed action; but the staff performed a documented
{                            evaluation which stated the objectivea of and reasons for the requested actions and the                                    :
c
~                            basis for invoking the compliance exception. See also 10 CFR 50.54(f) . A copy of this                                      ;
evaluation will be placed in the NRC Public Document Room,                                                                  ;
Federal Register Notification A notice of opportunity for public comment was not published in the ErderAIRegister                                        t because of the urgent nature of the generic letter. However, comments on the actions requested and the technical issues addressed by this generic letter may be sent to the U.S.
Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C.
4 20555-0001.
Paperwork Reduction Act Statement
,                            This generic letter contains infonnation collections that are subject to the Papermk                                        !
Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections were approved by the Office of Management and Budget, approval number 3150-0011, whid-
                            -- expires on July 31,1997.
The public reporting burden for this collection of information is estimated to average 300 hours per response, including the tirr.e for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. De U S. Nuclear. Regulatory Commission is seeking public comment on the potential impact of the collection of information contained in the-geniric letter and on the following issues:
(1)    Is the proposed collection of information neceswry for the proper performance of the functions of the NRC, including whether the information will have practical utility?                                ,
(2)    is the estimate of burden accurate?-
(3)    is there a way to enhance the quality, utility, and clarity of the inforraation to be                              ,
collected 7-                                                                                                      _
(4)    How can the burden of the collection of information be minimized, including the use of automated collection techniques?
i
  -.w,,,-w,e.    --,,w--,,    ,          ,, -,  .,.mn,,nw---,,n , ,---n+ , , - ~ ,,,,.a,- ,,m--  ,- ---,      .,_~,-,-a-,-nnnen.--,,,- -,-e-,,-'
 
5- ,
GL 96-06 Sept:mber 30,1996 Page 10 of 10 Send comments on any aspect of this collection cf information, including suggestions for s
reducing this burden, to the Information and Records Management Branch, T 6F33, U.S.
Nuclear Regulatory Commission, Washington, D.C. 20555-0001, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202 (3150 0011), Office of Management and Budget, Washington, D.C. 20503.-
The NRC may not conduct or sponsor, and a person is not required to respond to, a ,
j-    collection of information unless it displays a currently valid OMB control number.
l      if you have any questions about this matter, please cor. tact one of the technical contacts l      listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager, signed by B.K. Grimes Thomas T. Martin, Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contacts:  1. aura Duoes, NRR            James Tatum, NRR (301)415 2831                (301) 415 2805 Email: ladenrc. gov          Email: jettenrc. gov John Fair (301) 415-2759 Email: }rfenrc. gov Lead Project Manager: Beth Wetzel', NRR (301)415-1355 Email: bawenre. gov
 
==Attachment:==
List of Recently issued NRC Generic Letters
_ =
 
,    y RCV OY8hEl                          8-71-1) I 1:50 t      205 Ilt'6100*      Ng!:s t
  ,                                                                            NUREO-0933 A Prioritization of Generic Safety Issues                                                  .
Main Report and Supp.lements 1-12 is
      ^
it Bauit. R. Riggs, W. Mastead. J. Mttman Dividen of Satety Issue Rasolution Od5ee et Nuelear Regulatory Paesarela U.S. Nedear Rassistory r'a==1 tan                                                      '
Washington, DC 24555 6
4%ft).....
f N.,
_o
 
j                                                                                                                                                                Revision"5
  ,                                t 4
priorititetl and included in Section 3 and published in future                                                      -
l supplefsents to this report.
4 (4)-          Human Factors program Plan (HFPP) items identified for development
  ;                                                                  in MIREG 0985''' these items are covered in Section 4.
$                                                    (5)          Chernoby1 !ssues identified in NUREG-1851""; these issues are covered in section 5.
3 A comprehensive listing of all issues in the above five groups is given in
:                                                  Table !! which ir.cludes the fc110 wing information for each 'ssuet (1) the j                                                    Ntc person responsible for the prioritiration evalu'tioni (t) the lead NRC i                                                  office, divisten, and branch responsible for reviewing the prioritigation 1                                                  analysis and/or resolving the issue; (3) the priority ranking or status; '                                                                  ,
!                                                    c4) the latest version of the evaluation (5) the issuance date of the                                                                      !
;                                                    'atest version of the evaluationi and (4) the MPA nunber for those' issues
'                                                  that have been resolved and require licensee actions. A summary of the f                                                    number of issues in each category is shown in Table !!!. A cross-reference i
11sttng of reports prepared try the Office for Analysis and Evaluation of Operational Data (AE00) and their corresponding generic issues is provided
* in Table IV.
How the Werk is nana                                                          .
l'                                                  The work is done, in accordance with the criteria described below, by the responsible NRC Branch in consultation with others i'i the NRC with i                                                    kiewledge of the issues or expertise in the technical disciplines involved. In a number of instances, technical or cost infomation is obtained from industry and other outside sources. The Battelle Pacific l                                                                                                                                                                                          ,
;                                                  Northwest Laboratories (PNL , under a technical assistance contract, t
developed detailed- methods )to quantify safety benefits and costs and 2
provided safety-benefit analyses and cost information for many of, the
:                                                      1ssues. The responsible NRC Branch, with internal consultations as i-                                                    necessary, reviews and applies the PNL-supplied technical factors, in l                                                    conjunction with additional factors, in developing the priority rankings i
and recofnuendations.
!                                                    Systematic peer review cf each prioritization evaluation within the NRC contributes to the assurance that the analysis is complete and accurate and that the judgments are soundly based. This review is -sono in two
                                                  -stages. First, each analysis is reviewed by the NRC orilanizational unit or L                                                    units. whose area of responsibility or specia19:ed knowled                                                                    is substantially involved. Second, any comments made are thsn resolved,ge                                                    where l                                                    practical, and factored into the analysis, as appropriate. Upon completion 1-                                                    of peer review, the analysis is then finalized and prepared for approval l                                                    by the respenalble Office DiMstor. Once approved, it is p1ssed in the p0R
!                                                      and published in a future supplement to this report, after which, l
additional comments from the ACRs, the industry, and the public ars considered in any further reassessment of the issve's priority.
l priarity Catacariant Their Meanine and Pronesed Una l                                  s                  Feter priorisy eenMetye are woods H16H, MEDIUM, LOW, and oInor. They are intended for use in guiding allocation of NRC resources and scheduling of i
g                                                              NUREG-0g33            -
.                                          06/30/94 L                                                                                                                                                                                      v      .
  ...,e--      -- - - - - ,n,rw,,---,a,._w  ,-n. -w.-,--,.,m, .,m..        n.v-    ,    ,m,--~ ~ -  ,w- -
                                                                                                              , . . , --,-,,w.e-,-  r~~,-vw-,-,.-r-e,---,-,,,,---,.,-.,w-,-,,,y      -u,
 
      ,            (RCVSY8hEl                      I : > t 1:57 I              205 Ilt $100*                NElle 4 Revision 5
* a
,                                efforts to resolve the various issues, in conjunction with other pertinent factors such as 1) the nature, extent, and availability of manpower and material resource (s estimated to be requi.*edi (2) length of time needed to
,                                resolve; (3) conflicts in resource allccation and scheduling among items i                                of comparable priorityt (4) status of affected reactors; and (5) budget constraints.
1
:                                A tile priority ranking means that strong efforts to achieve the earliest practical nsolution are approtriate. This is because: (a) an important safety concern may be involved (though generally the conern is not severe enough to require prompt plar.t shutdown); or (b) the uncertainty of the
!                                safety assessment is unJsually large and en upper-bound risk assessment would indicate an important safety concern. All unresolved HIGH priority issues areyeriodically reviewed in accordance with the criteria stated in NUREG 0705 for possible designation as USIs. A 1111is defined as a matter affecting a number of nuclear power plants that poses important questions concerning the arcquacy of existing safety mquirements for which a final resolution has not yet been developed and that involves conditions not likely to be acceptable over the lifetime of the plants affected.5" In accordance with Section 210 of the Energy Reorganization Act of 1974, progress on the resolution of USIs is reported to Congress in each NRC
;                                Annual Report.
A          priority ranking means that no safety concern demanding high-pr or y attention is involved, but there is believed te be potential for
:                                safety. improvements or reductions in uncertainty of analysis that may be              /~
4                                substantial and worthwhile. Efforts at resolution should be planned,                  (.
perhaps over the ensuing years, but on a basis of not interfering with                    '
pursuit of HIGH-priority generic issues or other high-priority work.
4 A LQM priority ranking ameans that no safety concerns demanding at least MEDTIM-priority attention are involved and there is little or ro prospect of safety impMyements that are both substantial and worthwhile. When the prioritization process results in a t0W priority ranking for an issue, a> proval of this rankiry by the responsible office Director signifies that tis issue has been elin' nated from further pursuit. However, in accordance with SRM 871021A,"" the staff conducts a periodic review of existing LOW-priority GSIs to detemine whether there is any new information that would necessitate reassessment of the original prioritization evaluations.
TheggEcategorycoverspropcsedissuesthatarewithoutmeritorwhose significance is clearly negligible. Issues are also DROPPED froni further consideration if it is detemined that their safety concerns have been addressed in previously prioritized or res A sd issues. When the prioritization process results in a DROP priority ranking for an issue, a) proval of this ranking by the responsible Office Director signifies that tit issue has been eliminated from further pursuit.
An issue is considered resolved, indicated by NOTE 3 in Table II, when its resolution has resulted in either: (a the establishment of regulatory requirements or guidance (by Rule, SRP,), change, or equivalent); or (b) a documented authoritative decision that no change in requirements is warrarted. Priority rankings are not assigned to issues that have been                  '
resolved. However, in those cases where issues were resolved after having 06/30/94                                  10                            NUREG-0933
 
Revision 1 !
                    /*                                                                                                  l ISSUE 150: OVERDRFisVRIZATION OF CONTAINUNT_P_ENETRATIONS DESCRIP1103 l                                  Historical Bactaround                                              -
This -issue was identified ** by DSIR/RES and addressed the concern for overpressurization of contaitnent piping penetratier.s following a containment 1 solation and subsequent heat-up.                                                ,
1 Containment isolation at all nuclear powe plants ensures that redioactive .
materials are coatained if an accident or inadvertent release of such materials occurs. Isolation is provided for all piping systems that penetrate the containment. Double barriers are provided to ensure that no single failwo of an active component can result in a loss of this 1 solation function. Typically, this i                                  double barrier system is provided by isolation valves insioe and outside l                                  containment. Wnen containment- isolation is required because of an accident or l                                  inadvertent release of radioactive materials, these valves are closed to pre
* leakage cf radioactive materials to the environment.
~
lafety Sianificance                                                  -
Overpre!surization of the containment piping penetrations could potentially occur
~
during an accident -inve?ving- a significant increase in t;u containment temperature. This might occur when water that is trapped between the inner and i
outer containment isolation valves is heated and expands. Theoretically, heating a constart volume of water frem 100*F and 100 psia to 200*F we'.1d increase the pressure to 3000 psia. This pr:ssure increase ceuld fail the penetration or the isolation valves and could provide a direct flow path to the environment from the
,                                potentially contaminated containment atmosphere. The pressure increase is i                                  mitigated somewhat by the peretration itself expanding because of the tempersture
.                                in;;retse, as well as the possibility that the isolation' valves will not be lesk-tight and thus vill not pressurize fully.
~
Poss i bl e Solution E
A possib? solution to this issue was to prov1de a mechanism for preventing water i
from becouing trapped or for relievir.g
* ne pressure that could build up in the piping systems between the inner anu outer conta11 ment isolation valves, j                                  Licenseu would need to perforr, thermal and structural analyses of the penetration systems to determine whi .h penetrations, if any, were susceptible to such failure. A prsssure relief system would be nneded to prevent the pressure increase from failing the penetration. This preswe t slief systen could consist of the following:
(s)  Check valves inside the reactor building instead of the , inner
:ontainunt isolation valves. These valver. would ;revent water from s                                                becoming trapped between the two isolation valves out is only viable for penetrations with flow into containraent.
u6/30/95                                            3.150-1              NUREG-0933
    %.    ~
 
Revision 1 k or        r A method to provide                                tankpressure    relief, might be needed      such as to contain          a rupture blowdown
( dis (b)    safety valve. A story 11gald      or vapor that would be forced through the pressur equipment when the equipment is operated, PRIORITY DETERMINAI1QB k
Assumtions                                                                                              ident but This issus does not directly impact                                            the        l thatfor potential          might be a core      damage acc addresses a plant's          ability                            to        Thus, contain    the only concern radioactive        materia was thes a core damago accident.                                          failure  of  containment released during                                                                                  i    the effect probability of containikent failure resulting from                                    t failure,theassuming a
)Q        of the possible solution on the probability of contain
  .y      core damage event has cccurred, was evaluated,                                              lives of 28.8 and                      -
s  ~)                                                                                                                                            -
27.4 years, respectively, were                                              i l        affected by Ots isrve.                        I O were used as the reference rWR and BWR, respect ve y.
Fraauency Estimatn                                                                                    ;ontainment PRA" addressed failure of, or leakage                                  t through,For Oconee 3, the The Oconee 3                                                                                                    of a        _
penetrations as a potential failure scde of the containmen                                                        ed .  ('
probability of containment penetration leakage                                                                          1      .
as the base case value for Oconee 3.
Gulf 1 PRA" did not expli citly evaluate containment                                      t isolation penetration The Grand                                                                                              blic risk leakage because none of the                                                            accident d on WASH-1400            which sequences did perspective, However,              probability            tne Grand ofGulf  1 PRA was containment              base isolation        failure (#).
the WASH-1400" assess the conditionalTherefore, the base case value used in this anal analysis.                                                                                            the specific The prcbability of containment isolation failurecess                                              is dependent or failure          uponnm core damage sequence that occurs before cor                                                    tica  in Appendix containment isolation failure probability c?.nds                                        f      on the      prior suc of the various engin vred safeguards functions.tive                                              Basecprobability on in orma hted average V of WASH-1400," a value of 4 x 10-* was selected                                                            as  a representa for containment 1sciatton failure. This value tions                                                    in WASH-represented              the weig of the range of possible values based on the number cf observa 1400."                                                                                        l      for Oconee 3 The release categories asseM ated with                                                        t ories were isolation containment              not          fai ure are PWR-4 and PWR-5. Fct Grand Gulf 1, the iffected                                                            release ca eg" leakage, cxplicitly stated in  of NUREG/CR-2800" Oconee 3 that relatively              cr NUREG/CR-1659.high contaimn 1400"  and a P RA**'
attainable from f ailure of containnent isolation, woule prevenH,( gas exp containment building from potential everpressure caused                                                    e eventsby N.
Consideration incorporated into accident                              of thesecuences containment      involvincisolation containmentfail          ov frJREG-0933 3.150-2 06/30/95
      ~-
 
  . u Revision 1
    .-  caused by gas generation, as shown in Appendix B of NURN/CR-2800.'' A r.ew base case risk value was developed for all SWR - accident sequences that involve containment overpressure events. To accomplish this, the core desege sequences presented in NVREG/CR-2800" that could result in containment overpressure were modified to incorporate the base case containment isolation failure probability, rather than the containment overpressure: probability. This had the effect of creating a set of new accident sequences that included containment isolation-failure events.                                                                          -
The adjusted case values of the affected parkmeters were estimated by adding to the base case vaian-the prcbability of failure of the penetration system                                          .
(failure of one 01                            ore penetrations) that would arise from overpressurization.
A conservative appia:h was taken to develop a new containment laakage sequence of events that incorporated the potential for overpressurizatien. This sequence consisted of the following events: (1) containment isolation is successful; (2) water becomes trapped between inner and outer, isolation valves; (3) containment heating causes heating and expansion of the water between the isolation valves; and (4) the water expansion causes the penetration to fail, such' that a leakage path occurs between the containment atmosphere and the environment. Based on this seguence of events and using the rare event approximation, the additional probability of containment system failure was obtained from the product of the following terms:
N      =                            number of penetrations that are susceptible to overpressurization P[1]
probability water becomes trapoed= between isolaNn valve:;
P[2]  =                            probability inboard and outboard isolation valves are lenk-tight
                      =                            probability penetration overpressurizes to rupture, P(3) given that the penetration 1s -leak-tight and . full of water.
P[N)  =                            probability that the penetration fails in a manner that results in a leakage. path from the centaf uent                    ,
atmosphere to the ervironment,            given that the penetration ruptures due to overprassurization.
H depends on the types of penetrations and. isolation valves at each plant. Only liquid penetrations are susceptible to this- typa of failure and penetrations provided with check valves are nrr . To determine the Valde for N, the description of containment penetrations given in NSAC-60''' was assumed to represent both PWRs and BWRs. A total of 62 penetrations were listed, of which, 36 were provided with                                  s check valves or were not liquid-carrying lines and were not suseptible to this                                      6 cortainment failure sequence. Therefore, the value of N was 26.
ho inforttation was available to calculate P[3). Containment isolation valve closures are timed such that one valve closes slightly sooner than the other to prevent water frosi becoming trapped between the valves. Therefore, this event could be caused by failure of the containment isolation system control logic or 06/30/95                                                      3.150-3                                NUREG-0933
 
w    .
Revision 1 circuitry to fJnction as intended, or by failure of the valve to close when intsaded. P[1] was assigned.a value of 0.5.
In this analysis, P[2] was set equal to unity. Thus, no credit was taken for the fact that penetration everpressurization would not occur if one of the isolation valves were not leak-tignt. It was estimated that there was approximately a 30%
chance that one of these valves would not be lenk-tight.,"
To assist in estinating P(3), a simplit'ied engineering analysis was performed by 051R/RES to detertnine the stress and strain that a penetration would experience, assuming that it is leak-tight and fell of water. For the purposes of the analysis, a typical penetration was approximated as a 2" diameter,12" long cylinder fabricated of steel and having a yield point of 30 ksi. Assuming that the water was initially at 100*T and its temperature increased to 200'F as a result of an accident, it was calculated that the hoop stress would exceed the yield point. However, the volume of the cylinder would only have to increase 2.6%
to acconmodate the expansion of the water. This corresponded to a plastic strain l
of 1.3% in the diameter of the penetration, conservatively assuming no plastic strain in the axial direction. If the water were heated to 300*F, the diametric plastic strain needed to accommodate the exparsion of the water would be 3.8%,
again conservatively assuming no plastic strain in the axf al direction. These values of strain were far celow the values that would be expected to cause rupture of the penetration. Also, as e?pansion of the penetration volume occurs due to plastic deformation, the pressure of the trapped water decreases, further decreasing the likelihood of rupture. Using the above information, but considering that there is some probability that the material used to fabricate the penetration has an undetected #1aw, the vt.lue of P[3] was estimated to be 10".
(*
(.
For there to be a leak patt that satib.'ies the definition of P(4), there must be a failure inside and outside the containment. One possibility is that the penetration rupture " runs" past the containment vessel. The other possibilities involve failures of both containment isolation valves, or the containment penetration and one isolation valve. Such failures must be simultaneous since the failure of one component relieves tne pressure and eliminates the possibility of sequer,,ial failures. It was estimated that the value of P[4] was 0.1.
Based on the above estimates, the additional probability o.f containment system isolation failure was thS product of N, P(1), P[2), P[3], and P[4], and was approximately (26)(0,5)(1)(10")(0.1) = 1.3 x 10".
Conseognce Estimate Incorporating the values into the Oconee 34and Grand Gulf i PRAs resuit-4 in a potentir.1 public risk reduction of 1.3 x 10 man-ren/RY                            and 3 x 10** man-rem /Rf, respectively. Thus, the total potential pubite risk reduction was about 40 man-ren for all 134 affected p15nts.
Cost Estin ate Indost v_cos_t: To irrplement the poss. .. selution. licensees would be required to perform analyses to determine if certain penetrations were vulnerable to overpressuriution fo11cwing containment heat-up. These analyses were estimated                                g to require 4 man-weeks / plant. In addition, safety analyses and QA-related                                    s 06/30/95                              3.150-4                                                  NUREG-0933
  . s                              n                              .
 
w .;w Revision 1 activities would w needed because of the installation of hardware inside containment that requires considerable attention to- QA in' designing- the penetrations' pressure relief systems. Two man-weeks of labor per penetration were estimated for the design and safety analyses. The n.mber of vulnerable penstrations was- assumed to be 200 of the total penetrations without check valves, or ~ ' penetrations. Therefore, ' at 2 man-weeks / penetration, 14 man-weeks /p1 ant wuld be required.
Installing new hardware within containment at :perating plants (i.e., backfit) would require about 1 man-week of labor in radiation zones; plants under construction (La., forward 4it) would not require labor in radiation zones. The material costs were estimated to be 5900/ penetration and 56,300/ plant, including.
labor for pipefitters, welders, radiatinn monitoring staff, and instrument technicians. For forwerd-fit plants, the hardware costs remained the same, but .
labor costs would be reduced because personnel would not be working in radiation zones. Therefore a 50M redaction in labor requirements as estimated, i.e.,-0.5 man-week / plant.
Based on the above estimates, the total labor required was 19 man-weeks / plant for backfit plants and 18.5 man-weeks / plant for forward-fit plants. Therefore, at 32,270/ man-week, the estimated industry labor cost was $43,0CO/backfit-plant and 542,000/ forward-fit plant. With a tctal of 71 backfit plants and 63 fomard-fit plants, the total estiented industry implementation cost was 56.5M.
It was estimated that 28 man-heurs/RY would be require $ to conduct periodic
!.        (monthly) testing of the pressure relief system. At $2,270 v week, this cost l        . was 51,589/RY. Tor the 134 plants, the total estimated- Co . i JP operation and i        maintenance was 56M. Using a 5% discount rate, the pressnt worth of the recurring 1
costs associated with plant saintenance and operation was 53.3M.
M: It was estimated that 6 man 'nonths would be required for the staff to
        . develop acceptable methods, data, and acceptance criteria for licenstes to use when evaluating the vulnerabiljty of penetrations to the overpressure phenomenon analyzed in this issue. At s2,'470/ man-week, the total cost for this develope,nt was $54,000.
About'.Z man-weeks / plant were estimated for reviewing and evaluating licensee calculations of the stresses within the penetrations and for reviewing the design, safety analyses systems. At a cost of $4,and                        9 documentation 500/ plant, the total costforforthe penetration this effort was pressure estimatedrelief to.be $600,000.
After implementation, the NRC would have te inspect the operation and maintenance of the penetration isolation systems; I man-hour /RY was estimated as sufficient for NRC review of each system. Therefore, the annual lapor requi.rement was 0.18 man-week /RY for seven such systems. At 32,270/ man-week, the total cost for the inspection of the 134 affected plants was $1.5M. Assuming a 5% discount rate, this cost was $880,000.
Total Cost: The total estimated industry and NRC cost associated with the possible solutien to this issue was estimated to be 311.3M.
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s      .
Revision 1 F
3]pefimoact Assessment Based on an estimated public risk reduction of 40 man-res and a cost of 511.3M for the possible solution, the value/ impact score was given by:
S = 40 man-rem 511.3M
                                                                                      = 3.5 man-rem /$M Other Considerations (1)              For backfit plants, an estimated 5 man-hours / penetration of labor in a radiation zone would be required and, assuming 7 penetrations / plant, the total ORE would be 35 man-hours / plant. The dose rate was assumed to be 25 millires/ hour, which was representative of the dose rate inside containment during reactor shutdowns. The implementation dose was therefore kstimated to be about 0.9 man-rom / plant. For tl.a 71 backfit plants, the ORE was greater than the total averted public dose.
(2)              Routino testing and inspection of the penetration pressure relief systems were assumed to occur once every 30 days, similar to testing of the containment isolation valves. The testing was assumed to be performed by a 2-man team and to last about 10 minutis. Assuminij 7 penetrations to be i
tested, the total operation and maintenance dose was estimated to be about                        .
0.7 man-rem /RY.
(3)              The public risk reduction estimated for this issue was overestimated for several reasons. First, no credit was taken for the protection from overpressure that would be provided if one of the isolation valves were not leak-tight. Investigation of a previous containment issue indicated that there was approximately a 30% chance that one of the valves would not oe leak-tight.
(4)              The costs were estimated assuming that the containment isolation systems were located away from the containment building wall. This is not the case for many of the isottion valves which are located adjacent to the containment walls. This cssumption tended to minimize the costs because they would be clearly higher with modification cf the containment structure to accostnodate the pressure relief system. As a result, the cost estimates provided above were believed to be low.
CONCLUSION The estimated public risk associated with overpressurization of containment psaetrations was not significant. Based on the value/ impact assessment and the staff"s simplified engineering analysis, this issue was placed in the DROP category (See Aopendix C). In an RES evaluation,"" it was concluded that                                    '
consideration of a 20-year license renewal period did not change the priority of the issue.
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o    w Revision 1 EEf8 FEES
: 16. WASH-1400 (NUREG-75/014), " Reactor Safety study, An Assessment of Accident Risks in U.S. Comercial Nuclear Power Plants," U.S. Nuclear Regulatory Comissic,,1, October 1975.
54    NUREG/CR-1659, " Reactor Safety Stuey Methodology Applications Program,"
U.S. Naclear Regulatory Comission, (Vol.1) April 1981, (Vol. 2) May 1981, (Vol. 3) June 1982, (Vol. 4) November 1981.
: 64. NUREG/CR-2800,                  " Guidelines          for Nuclear Power Plant Safety Issue Prioritization                      Infomstion        Development," U.S. Nuclear Regulatory Coussission,  February 1983, (Supplement 1) May 1983, (Supplement 2)
December 1983 (Supplecent 3) September 1985. (Supplement 4) July 1986.
889. NSAC-60, "A Probabilistic Risk Assessment of Oconee Unit 3," Electric Power Research Institute, Ovnt 1984, 1330. Memorandum for T. King from W. Minners, "Overoressurization of Containment Penetrations," March 16, 1989.
1331. NUREG/CR-4220, "Reitability Analysis of Containment Isolation Systems."
U.S. Nuclear Regulatory Comissien, June 1985.                              ,
1554. Memorandum for W. Russell from E. Be:kjord, " License Renewal Implications of Generic Safety Issues (GSIs) Prioritized und/or Resolved Between October 1990 and March 1994," May 5, 1994.
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e 06/30/95                                                    3.150-7                    NUREG-o943 m                                -
 
o u QUESTIONS FROM 12/4/97 WORKSHOP Q1. Would you give more detail on the contents of the RAls (requests for additional information) that the licensees will be receiving?
A1,    Waterhammer and Two-Phase Flow issues in general, the RAls will be focused on the analytical methodology used, assumptions, inputs, conclusions reached and justification. The concept is pretty well captured in the waterhammer and two-phase flow slides that were presented at the workshop. Plants that do not credit containment fan coolers for accident mitigation will be asked to describe measures that have been taken to assure that the fan coolers will no: he used inadvertently following an accident.
Thermal Overoressure Issyg The staff has already issued a number of RAls conceming the thermal overpressure issue. Questions have frequently related to details cf the piping physical configuration or method of analysis on piping runs of interest. T'ie questions can be obtained from the docket files.
Q2.      What is the schedule for issuance of RAls?
A2. The staff hopes to issue RAls by the end of spring 1998.
Q3. Is the staffs review just based on RAls and licensee's submi'his or do you plan to perform site visits?
A3. The staffs review is primarily based on licensees'submittals and responses to RAls.
However, NRC staff may perform some sample audite to review licensees' '
implementation of corrective actions to GL 96-06.
Q4. Can a licensee use Appendix F beyond one outage?
B A4. Supplement 1 of GL 96-06 addresses this issue. The staff indicated that the use of AGME Code Appendix F _is accep?eble for interim operability determinations until permanent actions have been identified and approved by the NRC (as applicable) for resolving GL 96-06 issues. A licensee may request to use ASME Code Appendix F criteria for a permanent resolution of the thermal overpressurization issue by submitting a proposed FSAR amendment for staff review.
ATTACHMENT 3
 
2 QS. Why did the NRC single out the use of Appendix F in GL 91-18, and limit its use to only one cycle? @y are piping and pipe supports treated differently than otter operability determinations?
AS. The operability criteria for piping and supports was developed during the implementation of bulletins 79-02 and 79-14. Due to the significant impact of these bulletins at most facilities, the staff considered it necessary to develop a specific operability criteria that could be applied without prior staff review and approval. Appendix F (though a non-mandatory ASME, Section Ill, appendix) provides a thoroughly developed criteria that is acceptable to the staff in certain applications. Since this criteria involved a relaxation of the licensing basis criteria, the staff intended that licensing basis margins be promptly restored to the affected piping and supports. The use of Appendix F beyond one cycle is discussed with response to Q4 above.
: 06. Ara there any unsuitable reasons for extending schedules for implementing GL 96-06 corrective actionsi A6. Yes. As discussed in Supplement 1 to GL 96-06, specific actions that have been defined and are clearly needed should not be delayed without suitable justification (i.e.,
the GL supplement shoula not be used as an excuse to delay corrective actions).
Corrective actions should be prompt as required by Criterion XVI of 10 CFR Part 50 Appendix D, and the timeliness of completing corrective actions should be commensurate with the specific circumstances involved (e.g., safety significance, impact on equipment operability, complexities associated with compensatory measures t'lat are needed, and risk implications). See Revision 1 to GL 91-18 for additional guidance in this area I
l Q7. The industry is confused because we are not getting consistent guidance from the staff on USQs [ unresolved safety questions] and operability. GL 96-06 was issued as an urgent GL with no time for comment. However, now the staffis giving " blanket acceptance" for delays in schedule. Please, explain.
A7. This is really a two-part question. The first part deals with confusion about the              i relationship between USQs and equipment operability. GL 91-18 Revision 1, dated October 8,1997, was issued to resolve this problem and provides additional guidance in th!3 area.
The second part of this question asks why GL 96-06 was issued as an " urgent GL" with no time for public comment, when it appears that now the staff is giving " blanket acceptance" for scheduler delays. The GL was issued in response to design discrepancies that were identified by licensees that, in some esses, were serious enough to warrant action to shut down the plant. Cuisequently, the NRC felt it was important to inform licensees of the design discrepancies that had been identified and to obtain assurance that ar.y systems that were found to be subject to these discrepancies remained operable. l.icensees have now com;4ted their evaluations and have n                                                              -
 
      ,.          m 3-I 3
performed operability determinations as appropriate, and the situation is not as urgent
* l                            as it was when GL 96-06 was issued. While licenseas have found that corrsctive actions are needed to fully resolve the concems that_were discussed in GL 96-06,-
l                          . operability determinations are in place to assure the continued operability of safety-
,                            related equipment. Licensees are expected to take prompt corrective actions to fully
;                            resolve the concems expressed in GL 96-06, and the timeliness of completing corrective actions should be commensurate with the specific circumstances involved (see the response to QU, above). The intent of Supplement 1 to GL 96 06 is (in part) to assure that licensees have allowed themselves enough time for identifying and implementing I
corrective actions that are appropriate for the specific circumstances involved, while still i
being prompt. However, the staff has not given " blanket acceptance" of schedule
;                          - delays.
The staff recognizes that public comments are important and they typically result in a l                            better generic letter. While circumstan ,es sometimes warrant the issuunce of an
                          -* urgent
* generic letter (i.e., one that has not been released for public comment),- the preferred method is to solicit public comments and the staff will continue to implement -
this approach in accordance with agency procedures.
;                  Q8.      What is the time frame for issuance of Supplement 27
!                  A8. As discussed during the waterhammer and two-phase flow presentation, waterhammer vulnerabilities other than those discussed in GL 96-06 have been reported to the NRC.
Some examples include the Residual Heat Removal Service Water System, the Spent Fuel Pool Cooling System, and the Service Water System (during the station blackout-scenario). The staff is currently considerir.g issuance of a Supplement 2 to GL 96-06 to raquest that licensees review and evaluate system operating and design details to assure that systems important to safety are not vulnerable to waterhammer events. If the supplement is pursued, the staff would solicit public comments prior to its issuance and consequently, the supplement would most likely not be issued until sometime after August or September 1998.
Q9. . What impact will Supplement 2 have on licensees' ongoing efforts to resolve GL 96                              issues?
                  ~ A9.-    If issued, the staff does not expect Supplement 2 to have any impact on license 6s' ongoing efforts to resolve the issues discussed in GL 96-06. The intent of the supplement would be to focus licensee attention on other waterhammer concems that were not discussed in GL 96-06 (see the response to Q8, above), and ongoing efforts to address the GL 96-06 issues should not be affected.
Q10. Can Supplement 2 be a separate g?neric letter, so licensees can deal with them separately?
a
 
  .                          -+
4 A10. While it is possible to make the supplement a separate GL, the subject matter is closely related to *.he concems discussed in GL 96-06 and it would be most appropriate to issue it as a supplement to GL 96-06 to maintain continuity, However, the staff would not expect the supplement to impact ongoing actions being taken by licensees to address GL 96-06, and tne supplement could be handled as a sep; rate action by licensees (see the response to Q8 and Q9, above).
                                'Q11. When does the staff expect to close GL 96-06?
A11. The staff plans to issue safety eva!uations to closeout GL 96-06 for each plant. The current target date for final closeout of GL 96-06 is December 1999.
Q12. Regarding the options listed in your slide for potential fixes for the overpressurization                    <
                                        -Issue, are they for operability or permanent solutions? Has the staff found any of these options unacceptable?
A.12 The staff has issued RAls that address some of the proposals. Alti.ough, the staff has
            '                              not made a final determination on the adequacy of the proposed fixes, it has not
                                        ' identified any as unacceptable options.
                                -Q13. One fix we are considering is drilling a hole in a portion of the valve disk, which is the same fix that was used to address pressure locking. If we do that, how much testing must be dom to confirm this?
A13. The staff does not anticipate that testing or surveillance, in addition to that provided under existing maintenance and testing programs, will ba necessary to verify the :
adequacy of drilled holes for pressure relief.
Q14. What is NRC's position on subsection 7500 of the ASME code? -
A14.
Subsection 7000 of the ASME Code does not r>rovide criteria for Level D (faulted) loads.
The stress limits in subsection 3000 of the ASME Code are applicable to Level D loads.
Q15. Would the staff expound a little bit moie on the safety significance of this issue? With respect to GSI-150, is this a very safety significant item? Are the provisions of 50.109 met with regards to safety significance?
                                'A15.      As stated in GL 96-06, the staff considers the issue of thermally induced overpressure of piping a compliance backfit, under the provisions of 10 CFR 50.109. The staff still considers the issue a safety concem. However, as stated in GL 96-06, Supplement 1, licensees should consider all safety aspects when establishing correcuve action and schedules. Thus, if a proposed modification is st.fety negative, licensees should
 
  ..  -~
5 consider other means of addressing the underlying concerns, including other modification designs.
Q16. There are two criteria for the overpressurization issue (code complianc.e and L.). If you              ,
meet ASME code compliance you will meet containment requirements, but net vice versa. la it acceptable for a line to yield?
A16. The ASME Code Appendix F criteria allows the pipe to yield.
Q17. There's a line that comas into containment, but is isolated under accident conditions.
Provided that the licensee can show that the breach will not occur at the penetration and not cause c functionality problem with the system, is a breach acceptable?
A17. If a system performs no safety function, and if the failure of the piping will not irnpact any safety-related system or fission product barrier, then a licensee may be able to demonstrate the failure has no consequence.
Q18. Does the staff completely discount the theory that pipe expansion will take care of overpressurization?
A18. The EPRI tests clearly demonstrate that expansion of the pipe (due to the thermal expansion of the metal) does not fully relieve the pressure. The tests demonstrate that yielding of the pipe wall v/ill mitigate some of the pressure increase. However, the tests also demonstr te that the pipe wall yiciding will occur in locally thin sections of the pipe during the event. Depending on the piping ennfiguration, if locally thin areas exist, they could fait during the event.
Q19. Is it acceptable to use insulation on piping as a method of mitigating overpressurization?
A19. The use of insulation to mitigate the temperature rise in the fluid may be acceptable as long as the ir.sulation is properly cualified.
Q20. For systems subject to overprersurization from fluid thermal expansion, can licensees
,              use App. F criteria for interim operability justification (until final 96-06 resolution) for stresses caused by thermal expansion of pipe, beyond next RFO [ refueling outage)?
A20. Refer to the response to Q4.
Q21. EPRI model and testing is good, but does not resolve NRC concern that the issue is not safety (addressed under GSI-150) but compliance.
 
f,,    2,,,
          't O 6
A21. The EPRI test results may be useful in developing an analytical model to evaluate the thermal overpressure. Although the issue of containment bypass was addressed under GSI-150, the NRC stafiis concemed with situations involving system functionality (as described in GL 96-06) as well as compliance with the licensing basis, Q22. Will the NRC revise Reg Guide 1.141, " Containment Isolation for Fluid Systems," based on GL 96-067 When I checked last spring RG 1.141 has been out for comment for almost 20 years.
A22. There are no plans to update RG 1.141, if the staff should decide to issue guidance, or approve methodologies relating to protection against overpressurization of containment penetrations, it would be via a means of communication other than an update of RG 1.141.
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Latest revision as of 07:30, 10 December 2024

Summary of 971204 Meeting W/Nuclear Energy Institute Re Industry Workshop on GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design Basis Accident Conditions. Listed Documents Encl
ML20199G742
Person / Time
Issue date: 01/28/1998
From: Wetzel B
NRC (Affiliation Not Assigned)
To:
NRC (Affiliation Not Assigned)
References
GL-96-06, GL-96-6, NUDOCS 9802040336
Download: ML20199G742 (349)
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