ML20134P923

Summary of 961029 Industry Meeting W/Nei & Licensees Re Issues Pertaining to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions. List of Attendees & Handouts Encl

ML20134P923
Person / Time
Issue date:	11/22/1996
From:	Marsh L NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
GL-96-06, GL-96-6, NUDOCS 9611290380
Download: ML20134P923 (157)

Text

{{#Wiki_filter:. _ _ _. _ _ _ _ _ _ _ _ _... _ _. _ _ _ _ _ _ _ _... _ _ _ _ November 22. 1996 i t MEETING SPONSOR: NUCLEAR ENERGY INSTITUTE (NEI) j -

SUBJECT:

MEETING WITH NEI AND LICENSEES TO DISCUSS GENERIC LETTER i (GL) 96-06~ " ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENtINTEGRITYDURING-DESIGN-BASISACCIDENT 4 i CONDITIONS" On October 29, 1996, members of the NRC staff attended 'an industry meeting hosted by NEI to discuss issues pertaining to GL 96-06, " Assurance of i Equipment O Conditions,gerabilityandContainmentIntegrityDurin Design-Basis Accident in Dallas Texas. The NRC staff partici ated in the morning i portion of the meeting, which was publicly noticed. he afternoon portion of the meeting was open to industry participants only. Attachment I contains a i list of questions and answers that the NRC addressed during the meeting. ! 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' i regardingGL66-06. Attachment I contains questions that were received by NRC before the meeting for discussion during the meeting and questions that were i presented to the NRC staff during the meeting and the answers to both. The . attached clarifications do not 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 301-415-1355. Original signed by: Ledyard B. Marsh, Chief Plant Systems Branch Office of Nuclear Reactor Regulation Attachments: As stated (2) [ -b#gh] Jistrib t on: O Jocket T e (all BWR PUBLIC g/att 1 & 2)/PWR licensees w/att 1 & 2) y 0 g/ OGC ACRS I E-MALL (liaatt 1) R. Zimmerma/A. ThadaniR. Cooper,(JKR) ~. M' rag E. Jordan R. Wessman n RI G. Hubbard DRPW/DRPE E. Merschoff, RII J. Tatum \\\\ OPA J. Caldwell, RIII C. Hammer W. Dean, RI J. Dyer, RIV K. Manoly G. Tracy, RII L. B. Marsh J. Fair B. McCabe RIII C. Berlinger W. Long J.Mitchell,RIV R. Reedy DOCUMENT NAME: G:\\WPDOCS\\PRAIRIEjMTG. SUM see previous concurrence "o receive a cacy of th is dortament. Indicate n the box C=Crov w/ > attachment /en:losure E= Copy with ett ichment/ enclosure N = No copy OFFICE PM:PD31 sE LA:PD31 E SPSB nm SCSB EMEB 'D:PD3-1 CJamerson0' TMarsh W M CBerlinaer

• RWessman
• JHannon
• c WE BWetZ D

DATE O fy//96 // /2(/96! // [2-/96 11/21/96 11/22/96 11 / 22/96 FICIAL RECORD C0P 59 9611290380 961122 { PDR REVQP ER C

i MEETING SPONSOR: NUCLEAR ENERGY INSTITUTE (NEI) I

SUBJECT:

MEETING WITH NEI AND LICENSEES TO DISCUSS GENERIC LETTER (GLS 96-06 " ASSURANCE OF EQUIPMENT OPERABILITY AND i CON'AINMENiINTEGRITYDURING DESIGN-BASIS ACCIDENT CONDITIONS" On October 29, 1996, members of the NRC staff attended an industry meeting hosted by NEI to discuss issues pertaining to GL 96-06, " Assurance of Equipment O Conditions,prabilityandContainmentIntegrityDurin Design-Basis Accident in Dallas Texas. The NRC staff partici ated in the morning { portion of the meeting,, which was publicly noticed. he afternoon portion of i the meeting was open to industry participants only. Attachment I contains a list of questions and answers that the NRC addressed during the meeting. is the meeting handout, which includes overhead slides used during both the morning and afternoon sessions and a list of meeting 4 l_ 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 $6-06.the NRC staff addressed questions from the meeting participants' regarding GL Attachment I contains questions that were received by NRC before the meeting for discussion during the meeting and questions that were presented to the NRC staff during the meeting and the answers to both. The attached clarifications do not 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 301-415-1355. Ledyard B. Marsh, Chief Plant Systems Branch Office of Nuclear Reactor Regulation Attachments: As stated (2) listribJt'on: Jocket

i' e (all BWR PUBLIC (w/att 1 & 2)/PWR licensees w/att 1 & 2)

OGC ACRS PD 3-1 Rdg NEI E-MALL (w/att 1 . M'raglia/A. hadani R. Cooper (JKR) E. Jordan R. Wessman R. Ziinnerman RI G. Hubbard DRPW/DRPE E. Merschoff, RII J. Tatum OPA J. Caldwell, RIII C. Hammer W. Dean, RI J. Dyer, RIV K. Manoly G. Tracy, RII L. B. Marsh ' J. Fair B. McCabe RIII C. Berlinger W. Long J. Mitche 1, RIV ocebaEndo*ffh ?att t c e t C=C nt a =Co with attschsent/enetos c e N = No copy omCE PM:PD31 1E LA:PD31 E. SPSB SCS M 6-ENEB B/EBf,,I NME BWetZ SD CJamersond TMarsh CBerlinaer RWessman

• MN DATE ll /p//g6

// /21/96! / /96 // /gp /96 11 / 22/96 U /fA/96 0FFICIAL RECORD COPY

j l MEETING SPONS0k: NUCLEAR ENERGY INSTITUTE (NEI)

SUBJECT:

MEETING WITH NEI AND LICENSEES TO DISCUSS GENERIC LETTER (GLp96-06 " ASSURANCE OF EQUIPMENT OPERABILITY AND CON 'AIMENi INTEGRITY DURING DESIGN-BASIS ACCIDENT j CONDITIONS" ) On October 29, 1996, members of the NRC staff attended an industry meeting j hosted by NEI to discuss issues pertaining to GL 96-06 " Assurance of Conditions,prability and Containment Integrity Durin, Design-Basis Accident Equipment O in Dallas Texas. The NRC staff partici ated in the morning portion of the meeting,, which was publicly noticed. he afternoon portion of l the meeting was open to industry participants only. Attachment I contains a list of questions and answers that the NRC addressed during the meeting. is the meetina 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 06-06. Attachment I contains questions that were received by NRC before the meeting for discussion during the meeting and questions that were presented to the NRC staff during the meeting and the answers to both. The attached clarifications do not 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 301-415-1355. i Ledyard B. Marsh, Chief Plant Systems Branch Office of Nuclear Reactor Regulation Attachments: As stated (2)

listrib e on-

ocket N e ('all BWR PUBLIC (w/att 1 & 2)/PWR licensees w/att 1 & 2)

OGC ACRS PD 3-1 Rdg NEI f-1 (w/att I raglia hadani E. Jordan R. Wessman R. Zimmerma/A. R. Cooper (JKR) n RI G. Hubbard DRPW/DRPE E. Merschoff, RII J. Tatum GPA J. Caldwell, RIII. C. Hammer W. Dean, RI J. Dyer, RIV K. Manoly G. Tracy, RII L. B. Marsh J. Fair B. McCabe RIII C. Berlinger W. Long J.Mitchell,RIV J DOCUMENT NAME: G:\\WPDOCS\\ PRAIRIE \\MTG. SUM "o receive a cany of this dartment. Indica te in the box C=Ca ny m/, attachment /en:tosure E=ceov 1lifilhatt ht/enetostre N e No copy OFFICE PM:PD31 iE LA:PD31 E SPSB SCSB EMEBi_ D:PD3-1 CJamersonh TMarsh CBerlinaer RWeMI JHannon WWE BWetZ D ente ll/p//96 // /4/96[ / /96 / /96 U d/96 / /96 / 0FFICIAL RECOR) COPY l

aurg t n 'i UNITED STATES f j NUCLEAR REGULATORY COMMISSION 2 WASHINGTON. D.C. 206t50001 - %*****/ November 22, 1996 MEETING SPONSOR: NUCLEAR ENERGY 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 { CONDITIONS" t On October 29, 1996, members of the NRC staff attended an industry meeting hosted by NEI to discuss issues pertaining to GL 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident i 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. is the meeting handout, which includes overhead slides used j 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 that were presented to the NRC staff during the meeting and the answers to both. The j attached clarifications do nct change the scope or requested _ actions of j 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 301-415-1355. sL (( /L Ledyar B. Marsh, Chie'f Plant Systems Branch' Office of Nuclear Reactor Regulation Attachments: As stated (2) l i i

OVESTIONS AND ANSWERS FROM OCTOBER 29. 1996 MEETING QUESTIONS GIVEN TO NRC BEFORE THE MEETING: 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 single 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 cooling water system (safety-related and non-safety-related) that can lead to complicated waterhammer scenarios which must be considered and 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 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 not 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 technical 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 4 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. Q5. The GL uses the term " delayed sequencing of equipment," what is meant by this term? A5. The term " delayed sequencing of equipment" was used in the GL to inform licensees that they should not limit their evaluations to design-basis accident (e.g., LOCAs [ loss-of-coolant-accidents] or MSLBs [ main steamline breaks]) concurrent with a LOOP [ loss of offsite power]. The NRC believes that the potential for waterhammer in CFC [ containment fan cooler) cooling systems may exist due to plant-specific time sequencing of safety equipment during or following accidents and not just due to accidents 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 sequencing delays in starting the cooling system pumps, voiding may occur and waterhammer may occur when actual flow through the system is established. Q6. With regard to the overpressure piping issue, if a plant installs a new i relief valve, do they have to account for a new potential leak path? Can we weigh this against the probability of having an overpressure situation? I A6. Even though a new potential leakpath may be introduced by installing a new relief valve, it may be necessary to install the reliefs. A design-basis accident condition which then causes a thermal overpressurization should be considered in 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. Q7. 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 in determining operability. 08. 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 this " excessive conservatism?" A8. As stated in GL 96-06, the staff believes that the guidance in NUREG/CR-5220 may be useful in determining various types of waterhammer 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 waterhammer and determine the associated loads on a case-by-case basis.

i i i l ; i Q9. Discussion: l The NUREG goes on to give several examples of why the calculated loads are i. 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 i e). reduction in slug length due to steam breakthrough. One way to address these various reductions in waterhammer loads is to use the Joukowski equation with a sonic velocity of approximately half of the sonic velocity 5 in water with no air or other non-condensibles. The resulting waterhammer i 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? A9a'. As stated in this question, there may be actual phenomena associated with-l any actual event that may reduce the_ magnitude of actual loads compared to: i calculated values. However, there are substantial uncertainties involved [ in predicting these conditions, making it difficult to ensure that L favorable conditions will exist to ensure such reductions. Typically, i good engineering practice involves making assumptions that provide j reasonable assurance of a conservative assessment. Q9b. Has the NRC published any additional information on methods to more accurately calculate waterhamer loads since NUREG/CR-5520 (dated 10/88), especially low-temperature,_ low-pressure waterhammer loads? i-j A9b. The NRC has not issued any other technical guidance since the issuance of NUREG/CR-5220 regarding waterhammer loads. Q10. Some actual plant-specific 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.

Q11. In the GL'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 allcwed to temporarily overpressure and that the comment about code applicability is incorrect here. Can the NRC comment on this?- 1.

me 2 All. The staff does not agree with the assertion that the codes do not require any consideration of this overpressure. The ASME Code has different stress allowables depending on the service loading conditions. Licensees are expected to use the appropriate load combinations and stress limits. I QUESTIONS ADDRESSED TO NRC DURING MEETING: Ql. Is there an implementation date that NRC expects this to be completed by? A1. No, the NRC has not specified an implementation date by which licensees should complete any corrective actions. Licensees should follow GL 91-18 ["Information to Licensees Regarding Two NPC Inspection Manual Sections on Degraded and Nonconforming Conditions and on Operability") and 10 CFR Part 50, Appendix B, Criterion 16, for svaluation of the operability and implementation of corrective actions. H eliness of the fix should be commensurate with the safety significance of the problem. a 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, what'is the appropriate action? A2. Any possible problems with meeting the 120-day response should be promptly communicated with appropriate basis to the NRC project manager for that plant, who will discuss the particular scheduling problems with the l technical reviewers for GL 96-06. i Q3. (paraphrased from several questions on the same subject) If the plant is licensed for a LOOP coincident with a LOCA only, is it i mandatory to consider a LOOP after a LOCA? j J 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. 3 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 coastdown for infrequent normal operation i of fan coolers, such as surveillance? i A5. No, there is no need to 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 temperatures during normal operation to account for infrequent extreme weather conditions, such as high summer temperatures. Q6. Can valve leakage be credited as a basis for long term operability? 1

, 16. 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 of impact on other plant equipment performance. Q7. The NRC states today, " Don't go beyond your licensing basis," however, the 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) creates the potential to lose containment integrity, an evaluation would tm required. Q8. Do you have to consider the pressurization of non-safety loops inside containment and the effects of this pressurization on containment isolation? A8. Yes, they need to be considered to ensure that containment integrity is maintained. Additionally, consideration needs to be given to whether the potential failure would divert flow from safety equipment and prevent the safety equipment from performing its required safety function. Q9. 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 Appendix J program. A9. In the response, the licensee 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. Q10. Please expanJ 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 being considered is to allow the pressure to exceed the penetration piping design pressure 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. It is the intent of GL 96-06 that the integrity of affected system piping be assured, especially for the piping containment penetrations. Therefore, the structural integrity of the piping should 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 allow 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. However, licensees should ensure there are no post-accident situations when this system would be called on by the operators or by automatic features to function. Overstressing and potential pipe yielding may be acceptable for this system as long as containment integrity is assured in accordance with the plant design-basis. Q12. Can " Leak Before Break" be used to limit containment temperature after a DBA [ design-basis accident) on an interim basis in order to support 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" manner, 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 - a initiating event. Q13. Will a consistent position be articulated to the Regions from NRR relative to startup issues? A13. We will communicate the results ;f 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 communicate our positions.

And third, we routinely are in touch with the regions 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.

t Q14. With regard to heat input causing voiding, is it appropriate to take credit for a reduced heat transfer coefficient for steam condensing on the exterior of the tubes (transferring 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) l i A14. The staff recognizes that the uncertainty involved in predicting heat transfer coefficients may be significant. There should be reasonable l assurance that the assumed coefficients are conservative for predicting heat transfer. Further, the accuracy of the assumed values should be consistent with the level of detail of the analysis. That is, for bounding analyses that have little refinement of the various analysis parameters, more conservative coefficients should be assumed to account for overall uncertainty in the results. For analyses that use more accurate correlations based on test data for the various parameters, less conservative coefficients may be acceptable. i Q15. Is drilling holes in the disks of containment isolation valves (inboard disks, inboard valves) a possible solution? 1 A15. The drilling of holes in system components could provide adequate pressure relief for a thermal overpressure condition, provided the resulting degradation of the structural integrity and safety functions of the components are adequately addressed. Q16. NUREG 5220 states that combination of waterhammer and seismic loads should j not be required for pipe stress evaluations. Typically, FSARs include piping load combinations of LOCA and seismic that were designed primarily for reactor coolant loop piping. What is the NRC's position on the load combinations for service waterhammer? 2 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-induced loads should be combined with other loads (i.e., dead weight, thermal, and seismic) as required by the plant design basis. The plant-specific design-basis load combinations may consist of FSAR requirements, 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)? 1 A17. Air cavities of various configurations may be acceptable solutions to the. thermal overpressure issue. However, since air (and other gases) are soluble in water, it must be assured that the air cavity will actually exist when needed. 4

I w Q18. The Beaver Valley overpressurized. penetration was due to heating over a long time, but the GL alludes to heating over a short time in accident conditions. Is it your intent for us to answer for just accident conditions or both? l A18. Any potential thermal overpressure condition should be. evaluated, including any normal or accident conditions. 6 P l

,Q.. 4 ) i, t I I 4 1, J h 4 4 i NRC Generic Letter 96-06 Industry Meeting j October 29,1996 1 i 1 1 e e e a me m m em h - = a M M j .ms e e,. S M S j ] ^ G m m WW q M M t er -amma _ W 8 6 .endummun 4 e W g j> = wusamme M 9 MM y - asses _ _.. _7-m muun sew.e wamerere ere= r ...%d

1. 3*+

i-ena tste ep-+d+a* +A ed.- W 9.-sO4 W 5 41* % Q" F AW MM4 e w %% hi. J s 9 9 4 1 i

l

UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 2055S-0001 September 30,1996 NRC GENERIC LETTER 96-06: ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS i Addressees All holders of operating licenses for nuclear power reactors, except for those licenses that i have been amended to possession-only status. ) Purpose The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to (1) notify addressees about safety-significant issues that could affect containment 4 integrity and equipment operability during accident conditions, l (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. Backcround 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 containment air coolers may experience two-phase flow conditions during postulated LOCA and MSLB scenarios. The heat removal assumptions for design-basis accident scenarios were based on single-phase flow conditions. Corrective actions may be needed to satisfy system design and - operability requirements. 9609250096N

GL 96-06 September 30,1996 Page 2 of 10 (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 also lead to a breach of containment integrity via bypass leakage. Corrective actions may be needed to satisfy system operability requirements. The sections that follow contain additional background information about each of these issues. } Wa terham mer On February 13,1996, the Pacific Gas and Electric Company (PG&E, the licensee for Diablo Canyon Units 1 and 2), determineci 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 coi: current loss of offsite power (LOOP) or with a delayed sequencing of equipment. This condition was reported to the NRC in Licensee Event Report (LER) 1-96-005, dated Aprii 26,1996. l The Diablo Canyon units have five containment air coolers in each containment, these are typically used during normal plant operation to prevent excessive containment 4 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 containment to the respective unit's component cooling water system (a closed-loop system). PG&E reported that, during 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 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 coolers. PG&E determined that 4 the stagnant component cooling water in the containment 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 j margin to boiling. On June 20,1996, Westinghouse Electric Corporation issued Nuclear Safety Advisory Letter NSAL-96-003, " Containment Fan Cooler Operation During a Design Basis Accident," to alert its customers to the potential safety issue that was

GL 96-06 September 30,1996 Page 3 of 10 identified by PG&E (Westinghouse is the reactor vendor for the Diablo Canyon units). In NSAL-96-003, Westinghouse recommended that licensees review their containment cooling systems to determine 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 power plant) declared all four of the containment air coolers at the Haddam Neck plant 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-loop 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-96-003. The licensee's analysis predicted hydrodynamic loads in the service water system from waterhammer that exceeded piping and support structural limits. On August 12,1996, the staff issued Information Notice (IN) 96-45, " 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 Engineering and Licensing Activities at Haddam Neck-Connecticut Yankee." Among other things, the report identified an issue relative to two-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 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 design-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 and would ultimately affect the service water flow and the heat rm oval capabili-ties of the containment air coolers. Steam formation due to low pressure and high temperature in the service water system could reduce the service water flow rates through the containment air coolers to values below those needed to

] GL 96-06 i September 30,1996 Page 4 of 10 i 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 Service Electric and Gas Company (the licensee for Salem 1 i 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 l water isolation valves for the nonsafety-related turbine loads do not start to close until approximately 30 seconds into the emergency loading sequence, the service water system 4 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 I the containment air coolers could result in additional " flow restrictions" in the air cooler tubes, further decreasing the flow of service water through the containment air coolers as j 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. l Overpressurization 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 valve to the RHR heat exchanger on Unit 1, the motor-operated butterfly valve located inside the containment would not open (EN 30833). The licensee found that pressure in 1 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 section of piping was relieved by opening a drain valve, the remotely operated butterfly valve was opened without any trouble. The licensee concluded that pressure in the isolated section of piping increased 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). The primary component cooling water (PCCW) system at the Maine Yankee plant has a nonsafety-related 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 containment 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

0 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 pressure relief valve on each of the six containment fan cooler PCCW branch lines downstream of the supply check valves. On August 20,1996, the staff issued Information Notice (IN) 96-49, " Thermally Induced 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 discusses 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. Waterha m mer 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-ielated systems can be affected. During this accident scenario, tSe steam that is produced ir. the containment air coolers may accumulate in other parts of the cooling water system, restricting flow 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 both 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 safety-related and nonsafety-related containment air coolers may not require a loss of offsite power for this scenario to be valid.

GL 96-06 September 30,1996 Page 6 of 10 Two-Phase Flow in Safety-Related Piping and Cornponents Two-phase flow (i.e., both steam and liquid) in cooling water systems associated with the containment air coolers can significantly interfere with the ability of the containment air coolers to remove heat under design-basis accident conditions, and can interfere with the cooling of other safety-related components. These cooling water systems were designed assuming single-phase flow conditions (i.e., liquid only) and containment heat transfer analyses are based on this assumption. Two-phase flow is a much more complex situation to deal with analytically than single-phase flow and involves additional hydrodynamic loading considerations as well as flow, heat transfer, systems interaction and erosion considerations. Additionally, the steam that is formed during two-phase flow can accumulate in the cooling water system, restricting flow and resulting in a waterhammer as discussed above. Overpressurization of Isolated Piping Because of its thermal expansion, water heated while it is trapped in isolated piping sections is capable of producing extremely high pressures. This phenomenon is typically a design consideration. Piping design codes as far back as U.S.A. Standard (USAS) B31.1 (1967), have explicitly recognized the need to consider the effects of heating fluid that is trapped in an isolated section of piping. The potential for thermally induced expansion of fluid trapped in valve bonnets was one reason for issuing Generic Letter (GL) 95-07, " Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves." In addition, several information notices (ins) have been issued discussing the pressuriza-tion of water trapped in valve bonnets, including IN 95-14, " Susceptibility of Contain-i ment Sump Recirculation Gate Valves to Pressure Locking," IN 95-18, " Potential Pressure-Locking of Safety-Related Power-Operated Gate Valves," IN 95-30, " Susceptibility of LPCI and Core Spray Injection Valves to Pressure Locking," and IN 96-08," Thermally Induced Pressure Locking of a HPCI Gate Valve." The potential for systems to fail 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 post-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. I 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 accident. If relief valves are installed to prevent overpressure conditions, consideration must be given to the effects of stuck-open relief valves and associated environmental flooding and radiation hazards.

O GL 96-06 September 30,1996 Page 7 of 10 i 1-Reauested Action (s) Addressees are requested to determine: (1) if containment air cooler cooling water systems are susceptible to either waterhammer or two-phase flow conditions during postulated accident conditions; 1 I (2) if piping systems that penetrate the containment are susceptible to thermal expansion of fluid so that overpressurization of piping could occur. 4 in addition to the individual addressee *s postulated accident conditions, these items r { 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 NUREG/CR-5220, " Diagnosis of Condensation-induced Waterhammer," dated October 1988, informative and useful in evaluating potential waterhammer conditions, i .If systems are found to be susceptible to the conditions discussed in this generic letter, e addressees are expected to assess the operability of affecW systems and take corrective action as appropriate in accordan~ceMithlhe requirements stated in)GL 91-18, &CPR Part 50 Appendix B and as required by the plant Technical Specifications ( to Licensees Regarding Two NRC Inspection Manual Sections on Resolu ' of Degraded ~ and Nonconforming Conditions and on Operability," dated November 7,1991, contains guidance on the review of licensee operability determinations and licensee resolution of degraded and nonconforming conditions, .i l Requested Information Within 120 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 operability of affected systems and components as applicable, and corrective actions that were impTemented 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 specific circumstances involved. Reauired Response 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

GL 96-06 September 30,1996 Page 8 of 10 complete the requested actions, or choose not to submit the requested information, or are unable to satisfy the requested completion date, must describe in their response any alternative course of action that is proposed to be taken, including the basis for establishing the acceptability of the proposed alternative course of action and the basis for continued operability of affected systems 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(f). In addition, send a copy to the appropriate regional administrator. Backfit Discussion Title 10 of the Code of Federal Regulations (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 " system 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 accident and to maintain them at acceptably, low levels." Addition-ally, Criterion 44 of Appendix A specifies a " system to transfer heat from structures.. systems, and components important to safety. The system safety function shall be to transfer the combined heat load of these structures, systems and components under normal operating and accident 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 specified 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 accident 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 criteria which specify 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 potential for overpressurization raises concerns that these piping systems will not meet their 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

1 CL 96-05 September 30,1996 Page 9 of 10 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 code criteria, respectively; and that containment integrity and these safety-related piping systems and components will not be adversely affected by 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-dance with the provisions of 10 CFR 50.109 regarding compliance backfits, a full backfit analysis was not performed for this proposed action; but the staff performed a documented evaluation which stated the objectives of and reasons for the requested actions and the 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 Reelster Notification A notice of opportunity for public comment was not published in the Federal Reeister 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. J Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001. Paperwork Reduction Act Statement This generic letter contains information collections that are subject to the Paperwork 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, which expires on July 31,1997. The public reporting burden for this collection of information is estimated to average 300 hours per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. The U.S. Nuclear Regulatory Commission is seeking public comment on the potential impact of the collection of information contained in the generic letter and on the following issues: i (1) is the proposed collection of information necessary 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 information to be collected? (4) How can the burden of the collection of information be minimized, including the use of automated collection techniques?

O e GL 96-06 September 30,1996 Page 10 of 10 Send comments on any aspect of this collection of information, including suggestions for 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 collection of information unless it displays a currently valid OMB control number. If you have any questions about this matter, please contact one of the technical contacts 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: Laura Dudes, NRR James Tatum, NRR (301) 415-2831 (301) 415-2805 Emt.h: lad @nrc. gov Email: jet 1@nrc. gov John Fair (301)415-2759 Email: jrf@nrc. gov Lead Project Manager: Beth Wetzel, NRR (301)415-1355 Email: baw@nrc. gov

Attachment:

List of Recently issued NRC Generic Letters i .w

GL 96-06 September 30,1996 Page 10 of 10 Send comments on any aspect of this collection of in' ination, including suggestions for reducing this burden, to the Information and Rece- .lanagement Branch, T-6F33, U.S. Nuclear Regulatory Commission, Washington, DA 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 collection of information unless it displays a currently valid OMB control number, if you have any questions about this matter, please contact one of the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager. original signed by B.K. Grimes Thomas T. Martin, Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contacts: Laura Dudes, NRR James Tatum, NRR 1 (301) 415-2831 (301) 415-2805 Email: lad @nrc. gov Email: jet 1@nrc. gov John Fair (301) 415-2759 i Email: jrf@nrc. gov Lead Project Manager: Beth Wetzel, NRR (301)415-1355 Email: baw@nrc. gov

Attachment:

List of Recently Issued NRC Generic Letters Tech Editor has reviewed and concurred on 08/30/96

• SEE PREVIOUS CONCURRENCES DOCUMENT NAME: 96-06.GL To receive a copy of this document, indicate in the box: "C" = Copy w/o attachment / enclosure "E" = Copy w/ attachment / enclosure "N" = No copy OFFICE TECH OGC C:PECB:DR D:DRPM i

CONTS PM NAME LDudes' RHoeiling' AChaffee* TMartin J Fair' JTatum' DATE 08/21/ % 09/05/ % 09/06/ % 09/30/96 OFFICIAL RECORD COPY

o Attachment GL 96-06 September 30,1996 Page1 of1 LIST OF RECENTLY ISSUED GENERIC LETTERS Generic Date of Letter Subiect issuance issued To 96-05 PERIODIC VERIFICATION OF 09/18/ % ALL HOLDERS OF OLs l DESIGN-BASIS CAPABILITY (EXCEI'r THOSE LICENSES OF SAFETY-RELATED MOTOR-THAT HAVE BEEN AMENDED OPERATED VALVES TO POSSESSION-ONLY STATUS) OR cps FOR NPRs 96-04 BORAFLEX DEGRADATION IN 06/26/ % ALL HOLDERS OF OLs SPENT FUEL POOL STORAGE FOR NPRs RACKS 95-09, MONITORING AND TRAINING OF 04/05/% ALL U.S. NUCLEAR ) SUPP.1 SHIPPERS AND CARRIERS OF REGULATORY COMMISSION I RADIOACTIVE MATERIALS LICENSEES 96-03 RELOCATION OF THE PRESSURE 01/31/% ALL HOLDERS OF OLs TEMPERATURE LIMIT CURVES OR cps FOR NPRs l AND LOW TEMPERATURE OVER-l PRESSURE PROTECTION SYSTEM l LIMITS i 96-02 RECONSIDERATION OF NUCLEAR 01/31/% ALL HOLDERS OF 012 POWER PLANT SECURITY OR cps FOR NPRs REQUIREMENTS ASSOCIATED WITH AN INTERNAL THREAT 89-10, CONSIDERATION OF VALVE 01/24/ % ALL HOLDERS OF OLs i Supp.7 MISPOSITIONING IN (EXCEI'T THOSE LICENSES j PRESSURIZED-WATER THAT HAVE BEEN AMENDED REACTORS TO A POSSESSION ONLY STATUS) OR cps FOR NPRs 96-01 TESTING OF SAFETY-RELATED 01 /10/ % ALL HOLDERS OF OLs OR LOGIC CIRCUITS cps FOR NPRs L 95-10 RELOCATION OF SELECTED 12/15/95 ALL HOLDERS OF OLs OR TECHNICAL SPECIFICATIONS cps FOR NPRs REQUIREMENTS RELATED TO j 2 INSTRUMENTATION i OL = OPERATING LICENSE i CP = CONSTRUCTION PERMIT NPR = NUCLEAR POWER REACTORS

EPRl/NPG ) EPRI E#ods to Suppod Resolution of GL96-06 issues Avtar Singh Mati Meri!o industry Workshop on GL 96-06 Dallas, TX ( October 29,1996 1002VtA&M fW11/BS % EPRl/NPG TOPICS Status of EPRI efforts to support resolution of the Containment Fan Cooler system waterhammer issue Proposed technical plan to support resolution of the GL 96-06 issues. 1 SARA i.on,

EPRl/NPG Fan Cooler Performance During LOCA with Loss C ? Offsite Power LOCA & LOOP at time = 0 Pump Coasts down in less than 2 seconds Fan coasts down slowly to 25% flow after 20 seconds Power returned after approximately 20 seconds Pump restarts after approximately 30 seconds Cooling water temperature approximately 80 F SARA EPRl/NPG GOTHIC NODALIZATION aganne Pl"* C 1 rT-U hr e- -, l l 'l* l l 4 /A ! l .....y // v

== /j' ? %l c'e, j*ks.T.II.4Mf.'.4 l !l l 5.;:!fM l l ,. f. E-,;_H:......,. 7.....yM, ;! l l l l y i l l um l l*t,..........................Q _ _l + SARA

4 i s 1 i y EPRl/NPG i / 1 i GOTHIC RESULTS l Vaporization starts at around 12 seconds All the tubes and 3" risers are totally volded in l-about 21 seconds ) Boiling stops after tubes are voided Evidence of oscillations at tube exits l Horizontal headers are half full of steam j Water temperature in large risers is stratified SARA EPRl/NPG RETRAN Noding Diagram 958 5 5 I W I O I C 41 0 Typical af Sam tanks SAS S 904 4 803 3 3G2 2 tot 1 - Wortting Level finer SUB WF Retum Meeeer huet

? EPRl/NPG Pressure in Tubes in Top CFCU = ~' 80 ..D*? s 7 70 Tu e sect.n 1 C' w' @.,. Tube Sect.n 2 LA: +. .0 - - T ..cin3 L.. - - - r e sm.n. 1; 1 so .e .h[ l40 o' f = c s g30 l sl,, + ++ 3 >;y, + ..r : = ~i~- 10 ^ ry.N .el* % \\ ~' %,r f O O 5 10 15 20 25 30 SARA e EPRl/NPG RETRAN Results Vaporization starts in about 24 seconds a Only top bank is partially voided Boiling stops as tube voids Induced flow in lower banks and lower saturation temperature prevent voiding SARA 10CfWA&51&tt48 to

l 1 i EPRl/NPG Column Rejoining VOID (I22'D; r e ? p f WHAM t a 3 s 2 i e p ..... NA b .. i b OFF ON SARA EPRl/NPG Assessment of Waterhammer Mechanisms: Water Column Rejoining

• Velocity determined by pump / system characteristics
• Potentially most severe mechanism

. Mitigating factors - Due to boiloff and subsequent temperature stratification in the headers, bubble is not at vacuum - Potentially have a two phase water column which lessens the severity of any waterhammer SARA h

1 i i l i EPRl/NPG l Assessment of Waterhammer Mechanisms: Steam-propelled Water Slug l l

• Occurs downstream of the throttling valve
• Subcooled water slug accelerated by steam produced by flashing of hot water in the throttling valve
• Mitigating Factors

- Slug dissipation and breakup - Liquid shedding at' the pipe walls - Rayleigh-Taylorinstability SARA EPRl/NPG Conclusions of Waterhammer Assessment Waterhammer will probably occur during LOCA/ a LOOP scenario The severity of such waterhammers is mitigated by many factors Experimental data lacking on - details of voiding and mixing in headers - waterhammer in pipes with boiling i SARA

l l l l EPRl/NPG ) f Parameters impacting the issue and Resolution (cont'd) l System Operation - CFCU required during normal operation / accident - Sequencing of alternate power sources during LOOP - CFCU operation during LOCA plus LOOP - Configuration of system (valve lineups) SARA EPRl/NPG Generic versus Plant Specific Aspects 1 of the issue and Resolution

• Preliminary information indicates vast variability in design and operational characteristics

-Issue resolution is expected to be mostly plant specific = However, certain CFCU performance related aspects may be generic -Two phase flow and voiding -Initial and boundary conditions for potential waterhammer SARA

. _. - - ~. . - - -. - ~. - l l l l EPRl/NPG Analytical Elements of a Technical Plan 1 OBJECTIVES - Scoping analyses of severity of potential waterhammer loads - Approaches for determining realistic thermal-hydraulic conditions to evaluate potential waterhammer loads - Determination of realistic thermal hydraulic inputs for plant specific evaluations - Parametric sensitivity studies SARA 190r#Abu gerite m EPRl/NPG Analytical Tasks l

1. Demonstrate low probability of occurrence for combined LOCA plus LOOP based on available data j

l

2. Develop an approach to predict if and when the boiloff I

occurs in CFCU tubes - Containment pressure, temperature, steam /alr, fan speed / flow etc., timing of LOOP versus LOCA assumed to be specified based on existing plant analysis. l - Effects of CFCU multiple parallel tubes and banks l SARA

EPRl/NPG Summary and Conclusions Generic tools exist to analyze and assess the impact of issues identified in GL 96-06 Containment fan cooler cooling water systems have plant specific designs i Computer models describing unique aspects of the phenomena involved may require experimental validation Cost-effective resolution can be achieved via industry collaborative efforts SARA l i f

l Generic Letter 96-06 Industry Meeting

Purpose:

- NRC Clarification of GL Recuested Actions - E tility Information Sharing l - Lessons Learned from L tilities that have Performed Requested Ana..yses - Iden:ification of Poten:ial Generic Ac:ivi:ies l

i Generic Letter 96-06 Industry Meeting i l Meeting Format: i l - NRC Portion - Public Meeting i - Submit Question Cards for NRC Ques: ions - Utility /EPRI Portions - Ques: ions from Floor are Encouraged h' 2

Generic Letter 96-06 Industry Meeting 1 l Important Aspects l - Consistent Framework for Analytical Methoc s l - Operability Determinations l l l - Resolution Approaches and Schedules h' 3

NEI Activities Faci itate Industry Meeting Interface wit:1 NRC for Generic Activities, Concerns

• Coordinate Generic Activities as Appropriate h'

4

9 NUCLEAR ENERGY INSTITLTE ~ NRC GENERIC LETTER 96-06 INDUSTRY MEETING OCTOBER 29,1996 i NRC PERSPECTIVE ON ISSUANCE OF i GL 96-06, " ASSURANCE OF EQUIPMENT OPERABILITY AND i CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACClbENT CONDITIONS" P TAD MARSH, GEORGE HUBBARD GARY HAMMER, BETH WETZEL U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION l

PRESENTATION l + OVERVIEW OF GENERIC LETTER - BETH WETZEL + NRC EXPECTATIONS - GEORGE HUBBARD + INDUSTRY QUESTIONS +

SUMMARY

REMARKS - TAD MARSH 1 l

SCOPE OF GL %-06 (1) POTENTIAL WATERHAMMER IN THE COOLING I WATER SYSTEMS SERVING THE CONTAINMENT AIR COOLERS DURING A LOSS-OF-COOLANT ACCIDENT (LOCA) OR A MAIN STEAM LINE BREAK (MSLB) (2) POTENTIAL TWO-PHASE FLOW IN THE COOLING WATER SYSTEMS SERVING THE CONTAINMENT AIR COOLERS DURING POSTULATED LOCA AND MSLB SCENARIOS i (3) THERMALLY INDUCED OVERPRESSURIZATION OF ISOLATED PIPING 2

REASONS FOR ISSUANCE OF GENERIC LEITER i + THE PROBLEMS IDENTIFIED IN THE LEITER ARE COMPLIANCE ISSUES l + THESE PROBLEMS EXIST AT SEVERAL PLANTS AND ALL PLANTS ARE POTENTIALLY SUSCEPTIBLE i + TWO OPERATING PLANTS SHUT DOWN AND HARDWARE MODIFICATIONS HAVE BEEN MADE OR ARE PLANNED AT SOME PLANTS + NRC WAS DEALING WITH THESE ISSUES ON A CASE-BY-CASE BASIS 3 i

BACKGROUND WATERHAMMER + DIABLO CANYON, LICENSEE EVENT REPORT (LER) 1-%-005, APRIL 26,1996 .+ WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR SAFETY ADVISORY LETTER NSAL-96-003, JUNE 20,1996 + HADDAM NECK, PLANT SHUTDOWN, (EVENT NOTICE 30772), JULY 22,1996 4 i

BACKGROUND (CONT) TWO-PHASE FLOW IN CONTAINMENT COOLERS + HADDAM NECK, INSPECTION REPORT 50-213/96-201, "SPECIAL INSPECTION OF ENGINEERING AND LICENSING ACTIVITIES AT HADDAM NECK-CONNECTICUT YANKEE," JULY 1996 + WISCONSIN ELECTRIC POWER COMPANY i SUBMITTAL REGARDING TWO-PHASE FLOW IN THE SERVICE WATER SYSTEM DURING A DESIGN-BASIS LOCA AT POINT BEACH NUCLEAR PLANT, JULY 23, 1996 + PUBLIC SERVICE ELECTRIC AND GAS COMPANY (SALEM), (EVENT NOTICE 30900), AUGUST 20,1996 5 i r u

j t BACKGROUND (CONTP OVERPRESSURIZATION OFISOLATED PIPING i + DUQUESNE LIGHT COMPANY (BEAVER VALLEY), (EVENT NOTICE 30833), JULY 3,1996 + MAINE YANKEE ATOMIC POWER COMPANY (MAINE YANKEE), (EVENT NOTICE 30769), JULY 19,1996 + VERMONT YANKEE POWER CORPORATION (VERMONT YANKEE), (EVENT NOTICE 30786), JULY 25, 1996 l 6

t BACKGROUND (CONT) L NRC ACTIONS: + INFORMATION NOTICE 96-45, " POTENTIAL COMMON-MODE POST-ACCIDENT FAILURE OF CONTAINMENT COOLERS." AUGUST 12, 1996 l + INFORMATION NOTICE %-49, " THERMALLY INDUCED PRESSURIZATION OF NUCLEAR POWER l FACILITY PIPING." AUGUST 20,1996 + MEETING WITH THE LICENSEE FOR POINT BEACH, SARGENT & LUNDY AND FAUSKE & ASSOCIATES i + MEETING WITH THE LICENSEE FOR HADDAM NECK AND CREARE ENGINEERING STAFF + CONFERENCE CALLS WITH SEVERAL LICENSEES-i AND REGIONAL INSPECTORS t 7 i

t WATERHAMMER CONCERNS + HYDRODYNAMIC LOADS IMPOSED BY t WATERHAMMER CHALLENGES INTEGRITY AND FUNCTION OF THE CONTAINMENT AIR COOLERS CHALLENGES THE CONTAINMENT INTEGRITY i 8 i

-? TWO-PHASE FLOW CONCERNS + COOLING WATER SYSTEMS DESIGNED ASSUMING SINGLE-PHASE FLOW HEAT TRANSFER DYNAMIC LOADS POTENTIAL FOR WATERHAMMER IN OTHER i PARTS OF SYSTEM i 9

+ i i OVERPRESSURIZATION OF ISOLATED PIPING CONCERNS + WATER HEATED WHILE TRAPPED IN ISOLATED PIPING SECTIONS IS CAPABLE OF PRODUCING EXTREMELY HIGH PRESSURES i + PIPING DESIGN CODE USAS B31.1 AND VARIOUS ASME CODE EDITIONS ENDORSED BY 10 CFR 50.55a i RECOGNIZE THE NEED TO CONSIDER THE EFFECTS OF HEATING FLUID THAT IS TRAPPED IN AN ISOLATED SECTION OF PIPING i [ i k 10 I b

3 i l REQUESTED ACTIONS i ADDRESSEES ARE REQUESTED TO DETERMINE: + IF CONTAINMENT AIR COOLER COOLING WATER SYSTEMS ARE SUSCEPTIBLE TO EITHER i WATERHAMMER OR TWO-PHASE FLOW CONDITIONS DURING POSTULATED ACCIDENT CONDITIONS + IF PIPING SYSTEMS THAT PENETRATE THE CONTAINMENT ARE SUSCEPTIBLE TO THERMAL EXPANSION OF FLUID SO THAT t OVERPRESSURIZATION OF PIPING COULD OCCUR i 11

.~ l REQUESTED INFORMATION + WITHIN 120 DAYS OF THE DATE OF THIS GENERIC LETTER, ADDRESSEES ARE REQUESTED TO SUBMIT A WRITTEN

SUMMARY

REPORT STATING e ACTIONS TAKEN i CONCLUSIONS REACHED i e BASIS FOR CONTINUED OPERABILITY CORRECTIVE ACTIONS c 12

.l 5 REQUIRED RESPONSE + ADDRESSEES ARE REQUIRED TO SUBMIT A WRITTEN RESPONSE WITHIN 30 DAYS OF THE ISSUANCE OF THE GENERIC LETTER ADDRESSING THE FOLLOWING: e WHETHER OR NOT ACTIONS WILL BE COMPLETED l e WHETHER OR NOT REQUESTED INFORMATION WILL BE SUBMITTED i 13 e i

NRC EXPECTATIONS + WATERHAMMER i + TWO-PHASE FLOW I v + THERMALLY INDUCED OVERPRESSURIZATION OF ISOLATED PIPING h b I 14

WATERHAMMER + TRANSIENT CONDITION + ACCIDENT WITH CONCURRENT LOOP l + ACCIDENT WITH SEQUENCING DELAYS I k i 15 t - ?

i TWO PHASE FLOW + STEADY STATE CONDITION + SINGLE PHASE FLOW l + HEAT TRANSFER CAPABILITY / COOLING WATER FLOW + PIPING AND PIPE HANGER STRESS / LOADS i + LOCATION e FAN COOLERS e UPSTREAM OF THROTTLE VALVES i DOWNSTREAM OF THROTTLE VALVES I 16 i

4.has4.a m a.al &.4. 4as.4. kw a uh _ e e*. 4 5 m h THERMALLY INDUCED OVERPRESSURIZATION + CONTINUING CONDITION + HEAT INPUT + PIPING STRESS / LOADS e i h 17 1 i f I

w NEI QUESTIONS REGARDING GL96-06 Q.1: 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 that specifies 2-phase flow or single phase flow. Q.2: Is this a start-up issue? Q.3: BWRs 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 back to conclude that it doesn't apply? 1 =

QUESTIONS (CONT.) Q.4: What type of fouling factors should be assumed? ~ l Q.5: The GL uses the term " delayed sequencing of equipment," what is meant by this term? Q.6: With regard to the overpressure piping issue, if a plant installs a new relief valve, they have to account for a new potential leak path. Can we weigh this against the probability of having an overpressure situation? t I f

i I QUESTIONS (CONT.) t Q.7: Can the industry use PRA and risk based arguments to say that the system is still operable when they are doing their operability determinations? i Q.8: The NRC has referenced NUREG/CR-5520, " Diagnosis of Condensation-Induced Waterhammer," in previous discussions with utilities on the containment fan cooler (CFC) water hammer j evaluations. The NUREG states (in section 5.1.6, pg. 62) that calculated waterhammer loads easily estimate upper bounds, but l actual loads are usually lower by a factor from 2 to 10. l Does the NRC agree with this statement and would they allow t licensees to take some credit for this " excessive conservatism? l

OUESTIONS (CONT.) Q.9: The NUREG goes on the give several examples of why the calculated loads are so much higher than actual loads. The examples are a) cushioning by uncondensed steam or i non-condensible gas; b) compliance of the piping, hangers, and mounts; c) oblique impact; d) friction on the water slug; and e) reduction in slug length due to steam breakthrough. One way to address these various reductions in waterhammer loads is to use l 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 waterhammer loads would then be l more accurate relative to actual loads. j l Does the NRC agree with using reduced sonic velocities as a means j of obtaining more accurate results than can be obtained from the j standard waterhammer equations? i Has the NRC published any additional information on methods to l more accurately calculate waterhammer loads since NUREG/CR-5520 (dated 10/88), especially low-temperature, low-pressure waterhammer loads?

i i QUESTIONS (CONT.) P Q.10: Some actual plant specific 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? i u h i i t i I

h i QUESTIONS (CONT.) i Q.11: In the GL's backfit discussion for the overpressurization of piping, there are statements regarding meeting the ASME code. ASME i experts have said that there is nothing in the ASME code precluding this condition. Fluid systems are allowed to temporarily overpressurize and the comment about code i applicability is incorrect here. l Can the NRC comment on this? l l I f Y

i i Haddam Neck l 1 CAR Fan Two-Phase Flow Presentation i Nuclear Energy Institute- ) XRC GL 96-06 Forum Dallas, Texas October 29,1996 i Paul D. Mason l Supervisor, Design Engineering CYAPCO

SIMPLIFIED SERVICE WATER FLOW DIAGRAM l From Train B CAR n m Service Water Header Fan M CAR n I Fan M SW f Pump CAR n Fan M + r CAR n [ Adams Fan M Filter SFP SW Hx l l Pump CCW Hx EDG l +L Hx Turb Tra. A in Service Bldg I I I I I I I I I I Water m Header l Discharge Tunnel 5 I

-l f SERVICE WATER SYSTEM ELEVATIONS Ele 50 ft j Ele 39 ft Ele 38 ft tele 38 ft Throttle Valve i I L = 30 ft g Ele 32.3 ft L = 42 ft Ele 18 ft I Ele 16.8 ft Ele 16.8 ft } L = 240 ft [ i ( Ele 9.0 ft U f Ele 10.3 ft Ele O ft i Discharge Tunnel i i i

t i Augmented Service W ter Supply SELECTED MODIFICATION H FOR LOCA WITH AND WITHOUT A LOSS OF POWER 4 54 x r c CONTAINMENT _--i Vent EEE sI i ,1 r I i 1 i I >4 ? Service Water Supply i f h To Discharge Tunnel I L_____J I

i i CLOSED COOLING WATER SYSTEM MODIFICATION -l i l l \\ j m 2 M Heat i m M D scharge Service -> M = l M l >4 Closed Cooling Water Pumps 1 i

1 i -l ] i DECREASED CAR vent FAN EFFICIENCY l vent l , r i \\ J LW I vent , r s k , r s k i drain vent i , r I J L 1 7 M J L dra,n vent i m r J L 1 r -H J L dra.in vent H. 37 JL 1 r M J L drain f 1 r M: %A' J k 3 7 J L V drain V II >N g y Service Water Supply f To Discharge j Tunnel l I

f i SPARGER CONCEPT l t l l i l l l n r s k l U f -] 1 CONTAINMENT 3 E2 EJ EJ El l a-g, V u im m n i n-r' l l Service Water Supply L_____J To Discharge Tunnel i

-i vent -l SERVICE WATER BOOSTER t PUMP FOR CAR UNIT SUPPLY t [ l l t L2 l

g j

Augmented Service 4 Water Supply Booster Pump j k , 7 A i M w2 rm i 1 I 2 m

• • "j;"

[ ~ Sp Booster Pump pi + 1 A L I To Discharge Tunnel >4 l Bypass Line (Typical) l i

CAR UNIT COOLING COILS (NO FOULING) (75,000 CFM AIR FLOW,261F AIR TEMPERATURE) 120.0E+6

, _ _ _ ; m_ - _

~,= =: i' l...f -- __ fd ~ i in npa i ir j-h l __ '/ 1,,r. .+ 1

-E.,J. irg p_

100.0E+6 E =F~iEiH: _ s-y --] : =Z~. gg 90.0E+6 11E EF A '- s,- : C 80.0E+6 gg{~_._ p' yn f g===4 ,// .l- /7 3 E- =-E-~Z-70.0E+6 p' 3 i +E EliE5= - y 5 : =:=g===- ,sj M.-.[}:. -. - _.. g,' ,f

1.... ~

60.0E+6 g E ---~ _ _. = _ y 50.0E+6 i.ly! T~~-E_ = j' 7' I = s / p,' tu .. :_C ~ :.C: ~~~~ ~~ = 40.0E+6 7__ : __ -~ r

===:?= _ -_ n' ~ E== sE. :- f 30.0E+6 +--- / 5._ - -- - - / ~ 95F Service Water Supply Temperature h- ~~~~ -- E 'O OE+6 ~ Ei~. _ -O-33F Service Water Supply Temperature ~ f-- ] _{~- _=- _ -o-28F Service Water Supply Temperature @ 10.0E+6 __Z~. - - -.. 000.0E+0 L ~ ~-~~~~~ 0 100 200 300 400 500 600 700 800 900 1000 Service Water Flow Rate (gpm)

Loss of Normal Power Sequence c f T= 0 seconc s Loss of Normal Power (LNP) T=10 seconds EDG bus energized T=43 seconds 1st SW pump energized T=53 seconds 1st CAR fan energized T=60 seconds Non-essential loads isolated j i l

Basis for Selected Modifications .l at Haddam Neck + Keep full, keep pressurized water supply to CAR fans + Vent dissipates energy downstream of i throttle valves to atmosphere + Eliminates potential for excessive forces + Not excessively costly (versus alternatives) + Simple design concept \\ j

Closed Cooling System + Advantages l - eliminates two-phase flow - minimizes / eliminates fouling concern + Disadvantages - additional EDG loads i i - costly hardware changes (new or upgrade) - delays start of CAR fans due to new pumps - extended shutdown i t

Decreased CAR Fan Efficiency ] + Advantages t - lowers outlet temperature to eliminate boiling + Disadvantages 4 1 - heat removal capability is intentionally reduced 8 pass to 2 pass 3 of 5 coils in service i - negative impact on current safety analysis - costly hardware changes for heads, isol valves - more complex surveillance for fouling l - on-line coil realignment f l t

1 -t f @reare 1 CAR Fan Two-Phase Flow i i Presentation to Nuclear Energy Institute October 29,1996 i 4 l Dr. Paul H. Rothe Creare incorporated l I l Creare incorporated P.O. Box 71 Hanover, NH 03755 Mtg. 96-10-792 (603) 643-3800 i

Three Events i LNP Loss of Normal Power 1 LOCA Loss of Coolant Accident i LOCA /LNP Both, Contemporaneous Onset s \\ i 1 1 .\\ \\ \\ t i 1 i I I 2 MTG-96-10-792 I i l i,

... - -....... _.............. ~.. -,......... _. - -.... - - - - -..... - - - - - -. -.. - - - -.... -..... -... -. -.. I-! Three Sections AR FAN RMR RmRN PUMP _qg (CONTAINMENT) (AUX BUILDING) sss6 ' THROTRES CAR CCW FAN l @reare s re-ee-,o-7e2

1 r Five Phenomena i i Pump CAR Fan River Return i Section Section Section l / / /*

1. Column Separation f

2. Flashing No Yes j

i

3. Boil-Repressure

/* l i

4. Condensation-

/* ? Induced Waterhammer l

5. Slugging (Surge,

? ? Line-Clear,...) i I h 4 MTG-96-10-792 i

4 CY Decisions i

1. Fix (Not Analyze) River Return

2. Fix CAR Fan Section j

- Keep Full, Keep Pressure, Keep Flowing j (Avoid Slugging Potential)

3. Focus Creare Analysis / Test On LOCA with Power Available j

4. Treat LOCA/LNP as a Flow / Pressure Refinement j

h I I i Greare 8 t i Y

h Main Findings for CAR Fan Section Overview 20 Hours l e Throttle Limits, Suppresses Boiling e No Boiling in CAR Fan e No Flashing at Throttle inlet e No Flashing in Line j 1 Microview 100 Seconds e LOCA - Void Builds, Slowly Collapses e LOCA/LNP - Same, with Accumulator i @reare e

Variables for Transient Analysis i l l P h f 1 i p V Throttle Valve j P 4x5O ^ g_ O P,T y y y l Op CAR Fan T,P f x x O P p l x OB - 20 Foot Tall i O,T,P Riser i g g L Y - 20 Foot Tall (40 Foot Long) l Downcomer (30 Foot Long) l - 240 Foot Run f i f 7 MTG-96-10-792 i i

I ll f f l:!!; ,[ fl !! ; i, >tI>!l!:I:!r?;f l iip 1 f g D eW 0e oM eE ue &nC e 9 e sI E_ b<Oi a s ) F ~ )F ( 2 ( tn tn e e m n m n n n ia i ta $ t E n n o o C C n in _s i 5 r e e r u u t ta a r ~ r e ep s m8 ~ p m e e T T m ma 8 a 8 e e t t S S o, o. t. n ~ g 5~ tL _2-a 8 s 8 ~ $1 $ .n + g g mw $ ?"9yWN n

I Water Temparature and Subcooling at Valva Inist V and CAR Fan Outlet X: Thru One Day Post-LOCA 100 One One 100 One One sec Hour Day sec Hour Day e 40 300 3 i X 30 c c S y v Sm m e 2 20 2 e g E 150 F-10 100 o so 1e+0 1e+1 1e+2 te+3 1e+4 te+5 le+1 1e+2 1e+3 1e<4 te+ Time (sec) Time (sec) l l r i t @reare "'a-**-' a-'*2 I i

b i Transient Motion of Interfaces . l t Line 151, LOCA Only l i 1400 350 ~ <--- Total Pipe Length 1200 l 300 CAR Fan Exit to Valve 2 <--- Start of Hot Water [ $ 250 Riser Reaches Valve

1000 a

E Steam Void cn Starts to 7 800 S l h 200 Collapse y _O 150 / \\ 2 600 Hot Water !2 k I Column Grows n. 100 gth Time 400 \\ [ 200 E N 50 Start to ,/ Dow mer Make Steam N 0 O L 0 10 20 30 40 50 60 70 80 i Time (sec) Position of Leading Edge of Hot Water Column - - Flow Rate Through Service Water Valve - - Position of Leading Edge of Steam Column t @reare

Margins -l t i i i

1. Maximum CAR Fan Heat Transfer i

- No Degradation l j 4 - No Thermal inertia

2. Multiphase Flow Regime

- Segregated Flow Treated Soft Landing If Bubbly Flow

3. Loss of Downcomer Head Neglected i

l i e f I h 11 MTG-96-10-792

Throttle Behavior i l e Various Throttles - (Orifice, Control Valve, Various Valve Types) e Various Configurations e " Specifications" - Pass Desired Cold Flow at Low Inlet Pressure - Increase Pressure Drop with Hot Water at inlet - Withstand Two-Phase inlet, Choking, Cavitation i A { I 12 MTG-96-10-792 i

Creare Throttle Facility Test i @ FLOW RATE @ LEVEL ( @ PRESSURE { @ TEMPERATURE I 3 12" HEADER x / A.R _, V5 I SUPPLY w T NK TAN FULL SIZE ir HOT WATER T T LOW 250T c' METER STEAM EXHAUST f fV3 X V4 y,M o + l g XV1 P STEAM '^ ER SPARGER ' vl R E,ER lf TANK I \\ { PUMP ~ STEAM Y BOILER i @reare m l

Throttle Test Findings i i (For CY Valve and Configuration) i e Hot Water Flow Rate is About Half of Cold Water Flow Rate at Same Pressure Drop (Tested Full Scale, Full Flow, Prototype Pressure). e Calculated Flow Rate Varies Only Slightly j (About 20%) as Plant Undergoes Thermal / Pressure Transient j i f @reare l '4 To-ee-,o-7e2

-I Conclusions .i t J With Power Available i (or With LNP and Keep Flow / Pressure) l Loads Will Be Minimal in CAR Fan Section During LOCA i - With Bubbly Flow, Soft Landing l - With Column Flow, Void Just Grows, Then Slowly Collapses i t - Void is Quantified To Be But a Fraction of the Pipe Volume, and is Limited Inherently l I l I @reare e t

i GENERIC LETTER 96-06 ISSUES " ASSURANCE OF EOUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN BASIS ACCIDENT CONDITIONS" PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON UNITS 1 & 2 October 29,1996 l Jorge R. del Mazo l

AGENDA i . Description of the Component Cooling Water (CCW) system Description of Scenarios Engineering Analyses Heat Transfer Calculations Water Hanuner Calculations Pipe Break Size Analysis Revised Heat Transfer Calculations System Fix Thermodynamic Effects

CCW SYSTEM SCHEMATIC "B" D k k SURGE TANK T T CFCUS EL '6t*-O' sc w [ } EL 171*-0* sr CCW PUMP HX Sl pp COOLERS COOLERS COOLERS ( ) CCW PUMPS e '~ CCW HXS l D

• B*

HEADER ASW - A P' EL 74* RECIPROCATING CP COOLER-RCP OIL COOLERS 2 3 9 ~ i*C" HEADER SEAL WATER HX ASW-o -~ LETDOWN HX 1. 3 ('. d '. SPENT FUEL Pli MX h a a 4. - EXCES LETDOWN HX i i i '~ x

• A*

HEADER D CFCUS El I61 -0* CCW CCP Pp S y' 'p " PUPF COOLERS COOLERS i. COOLERS RHR HX s-s. "C" "A" EL I4! -0* "B" s-s- ,, A " s-s_ "C" u -.z m-a m m m .- mJ

-_ww--w -.ww,.-w w-_ - - - - - -------;-'=4-mhem-s.mL4aaew-.mmwe sma-.m - rssa,,s.a amm -m.- ^ -'Am, f 1 H

Q 4.,

l [tT f; 1 4 lt i!: b, !J1., \\1( ; (i;< .w~j a w #.. [ l i~ " % t:(,tj ( yyg ;:f. ij ' i l l l gNNrf.y::.]g .g .-,= ~#' b*.d a 5 TOY. i rg jh 1 r 3 = i b, o.

qu's.

's, ,u,# ' -f f '

b. g.,

. -{ d b l ,rr...

• w N

yse I

1. • .I W

3 I p , ~. g',A. I I r l 6 DCPP UNITS HAVE 6 UNITS HIGH AND TWO BANKS DEEP 1 i 8

a PIPE CAP f \\ 3"dTYP\\ l l 12 g - /, a c .i l co n. i e d v i e* I i 6"p l n N0ZZLE TUBES V 3"d HEADER TYP ELEV 140' (7g% p ) \\ \\ \\ \\ N N s \\ \\ s s N N N ~, CFCU COIL (TYPICAL)

i SCENARIO 1 l LOCA/ LOOP (NOMINAL, IN SECONDS) 0 LOCA + LOOP OCCURS (t = 0) T 2 SI SIGNAL GENERATED EDG'S START I M E 4 12 VITAL EQMT LOADING SEQ. STARTS FLASHING IN THE CFCUs STARTS 12.5 9 22 CFCU FANS RE-ENERGlZED FIRST CCW PUMP ENERGlZED 26 30 LAST TWO CCW PUMPS ENERGlZED 4

0 4 5 3 sn o it 0 idn 3 oC tne mn ia tn r o e C 52 w tn o e p dic o c n A ) r s h e d t d n i n o w u N c 0 e e e t 2 s m a ( T im e i m ts i E T sv n w w o o D lF wo 5 l g 1 U S C n F a F C 0 1 5 \\ 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 0 9 8 7 6 5 4 3 2 1 1 1 =t

l l HEAT TRANSFER CALCULATIONS . Initially done by PG&E Officially done by Westinghouse Westinghouse program does not do water hammer e WGOTHIC code used Heat transfer coefficients Calculation backed by test data Enough heat and time to void the whole coil during LOCA Calculation independently done by Numerical Applications Inc. W calculation independently verified by Fauske and Associates Inc. MSLB is no problem because steam partial pressure

WATER-HAMMER ANALYSES . Bechtel Hydraulic Systems Transient Analysis (HSTA) code used . Program does not do heat transfer Output gives pressures and forces Stress analysis done separately Difficulty in establishing boundary conditions More voiding is less conservative Fluid dynamic effects affect amount of voiding (surge tank lines) ) Equipment cross sectional area not constant Steam bubble vapor pressure difficult to estimate Resultant water hammer forces are very high Water hammer forces affected by assumed steam void vapor pressure Qualitative effects on steam void vapor pressure (water temperature) Qualitative effects on water hammer due to changing vapor pressure

i 1 l i REDUCED BREAK SIZE LOCA . Evaluated all postulated break locations Pipe displacements based on pipe rupture restraint gaps \\ Pressurizer surge line break 144 squared inch pipe break analyzed 4 I Longitudinal break analyzed and not required to be postulated No boiling occurs before pump restan for concurrent LOOP i i i 4 4

PRESSURIZATION SYSTEM = Pressurizes CCW surge tank Surge tank design pressure is 30 psig Pressure set at 20 psig Pressurizes with Nitrogen with bottles and air back up Prevents boiling by raising the saturation temperature in the coil i i I 1

s SCENARIO 2 DOUBLE SEQUENCING (NOMINAL, IN SECONDS) I O SI SIGNALINITIATED 1 SLOW TRANSFER VITAL 4 Kv BUSES TO 230 Kv T I I M E i i i ) 30 UNIT TRIP TRANSFERS 12 Kv BUSES TO 230 Ky l l 50 SLUR INITIATES VITAL 4 Kv BUSES SHEDDING 52 DG BREAKER CLOSES, SI SEQUENCING LOADS FLASHING IN THE CFCUs STARTS 54 CFCU FANS ENERGlZED 62 66 FIRST CCW PUMP ENERGlZED LAST TWO CCW PUMPS ENERGlZED 70 ,r

n t i i THERMODYNAMIC EFFECTS l Done by EPRI e Analysis not completed e RETRAN and GOTHIC codes used e Thermodynamic effects Water temperature gradient prior to pump restart

Nuclear Energy Institute Industry Meeting to Discuss NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Condition" Dallas, Texas October 29,1996 1 Presenters: Steven Greco Project Engineer Wisconsin Electric Power Company Michael Flynn Project Manager Sargent & Lundy Robert E. Henry, Ph.D. Senior Vice President Fauske and Associates,Inc. Title of Presentation: Operability Assessment for Transient Conditions in the Point Beach Nuclear Plant Senice Water System During a Coincidental Loss of Coolant Accident and Loss of Offsite Power Outline:

1. Introduction - Steve Greco i

2. Transient Heat Transfer Analysis and Piping /Suppon/ Coil Stress Analysis - Mike Flynn i

3. Water Hammer Load Bounding Analysis Using Experimental and Analytical Methods - Robert Henry I

I i l l

-l Wisconsin Electric Power Company .l Point Beach Nuclear Plant .l Presentation Of Bounding Analysis Related To Westinghouse NSAL-96-003 and Generic Letter 96-03 Presented to: Nuclear Energy Institute t Dallas, Texas l October 29,1993

1 Wisconsin Electric Power Company .j Point Beach Nuclear Plant .J e Presentation Of Bounding Analysis Related To Westinghouse NSAL-96-003 ) c and Generic Letter 96-06 e Operability assessment for transient l conditions in the Point Beach Nuclear Plant Service Water system during a LOCA and LOC >P l l

.i .i Wisconsin Electric - Point Beach Presenters: Steven Greco, Project Engineer - WE Michael Flynn, Project Manager - S & L Robert Henry, Ph.D., Sr. Vice President - Fauske and Associates, Inc. l t

Wisconsin Electric - Point Beach Presentation Outline: h Introduction i Transient heat transfer and stress analysis Water hammer load bounding analysis l l l l i

9 Issue Review L Westinghouse NSAL-96-003 issued i - Potential for water hammer induced failure -LOCA & LOOP w/ delayed SR equip. restart l - Plant specific analysis required i I I

Issue Review (continued,; l WE review found applicability to PBNP ? -Review supported by FAI and S&L I -One-hour event report to NRC on August 1 -Initial operability decision l i l I ( l F I l

.i i ~I Post 1 hr report NRC communication e August 2 conference call: discuss detail e August 8 conference call: analysis results e August 15 meeting with NRR in Washington f e September 9 submittal to the NRC

Condition Description i e LOCA & LOOP Occur I e SW Pumps and CFC Fans trip e Column separation occurs in CFC piping i e Resident water in CFC coolers boils I i i i i o

i Condition Description (cont.} [ e EDGs energize bus in 10 seconds i e SW pumps restart at 25 - 35 seconds e Water hammer postulated on refill e Piping integrity challenged i

yt$ gm$C>:o,OagmmOC zoM p n. ps in p s s mo g n n 'C ig ut iF P S ioC B l e w Wo B ) SlF Se I n ( p t o ra s ) ) a VG 0 s V' t a sO p A gM mM uC pF iC S ~ aO W u~ ~ ~ Ob

• aU
• o

~ I u_ a r, ^ ~ gm ~ ~ ~ ~ 9m@ u ~ gu

• a i

u~ Lm ~ u~ ~ ~ ~ ~ ~ 5Om M5 mCE3-la d r r r t t t s sn a a C a a I) o or E S D u p leg s o t t t WS)N WS)B WS)F / / ( p m 's sS e L i l ro8 m k iuC IDS eU 2 m 2 m I I I es r e I sp SsS2 SsS2 SsS 2 t d PF p( p(3 p(3 loo a t 3 2 m a S C i P P W A Dk / u P u t u S aC i P P P IsL r c P W tSO L S =E,t=4s3 2= l d,. 1 ,l l! l i l l ll!1 '1ll,

1l!l 1ll

Y . * ~., i / PBNP CONTAINMENT FAN COOLER L AYOUT i. s. \\ l / \\ [ j j CFC EL. 84' Ilx I I EL. 67.5' l' o o i CCW EL. 55.5' 1 i ilX l .- I - z. W-20" Sw 25! RETURN.HDR. 3 I zl p 0, [ EL. 34.3' i l I ...] 1 C><0 CFC SW p TilROTTLE VALVES g, X EL. 8.3, CFC M0V'S. I 2907. 2908 i -, I I / / ![ Sw Puur UNil 2 UNIT I SW DISCil. { SW DISCll.

a '!Il4 I3i1 g ;i!4, .} 4 ~ ww~b 1 1 2 2 3 3 4 4 5 5 6 6 7 7 8 8 0 0 5 0 5 0 5 0 5 0 5 0 5 0 5 0 5 ~ M L 1 H X T - O 1 5 C ~ 3 . I : I - ! :i ! :I .. I: I ..I:1:I I. i 5 .i.I

[

(,! / /. / 8 2 A e' z" 1 S1 H HH cLo o X EX E-o 1 T1 - 5 5 1A U A 2 8 2'" s e i. !. i! I. ' -.i :i., N.1:l ;I"g 4i i 3 N. E

1

( .i.I \\. J I

e w$g 2z L' C 5 1 1 2 2 3 3 4 4 s s 6 6 T 7 8 E 0 5 0 5 0 5 0 5 o s 0 5 O 5 0 S C)< = ] 1 H X = 1 - 0 N5 1 8 i 1 1 1 -, ':i I a

ji

-I . ! : T (.! I :i y ',/ / 8, S1 HH 2 EX 2' E - T1 1 5 4D H 9_ N7o X ~ o - 1 5 g 0 s s 5 u 1 t . i I : I i i

"83 EWzl

T

l :

k ' .iI \\ 'i .t t

I -l Initial Operability Assessment i e Configuration limits water hammer impact e Short duration loading minimizes effect i i k r

Additional Work Completed e Detailed bounding analysis i -Thermal-Hydraulic Analysis -Piping Stress Analysis -Time-Response Analysis l e Experimental analysis -More realistic water hammer loading i -Conservatism in bounding analysis

.i Results of Bounding Analysis .l t i e Stresses within operability criteria 1 -But exceed design code allowables i e Heat removal function restored per FSAR o CFC system remains operable

&f Results of Experimental Analysis i e Scale model simulates CFC i confiauration l I e Minimal water hammer loading i e Confirms bounding analysis t conservatisin i

a i ~ Proposed Resolutions e Validate the experimental analysis j 1 e Modify piping / supports e SW Pumps with diesel drivers e Isolate & pressurize CFC circuit j i e Modify during 1997 outages

INDUSTRY CONTACTS i e Connecticut Yankee i -Similar scenario to PBNP 9 -6" Piping /non-seismic outside containment - Longer piping runs / deeper loop seals ) -Creare is consultant i eCP&L l -Similar scenario to PBNP -Reached similar conclusions to PBNP l i - Bechtel is consultant

s INDUSTRY CONTACTS (cont.) e Prairie Island i -Similar scenario to PBNP -Diesel driven SW pumps -Fauske and Associates is consultant l t e EPRI -Establishing a Tailored Collaboration Project -NEl communicating with the NRC -WE will participate

_m .m_. i TC-1 p, 3.4 m l W11 O N ~T 3 N* So6enoid Operated Bat Vafves -- TC-2 4 59m

• • a Evacuated N

Receiver 2 Vessel i --TC-3 Manual Basvake 3g ~70 psig P. P - -28.6 "Hg onhee Y TC-8 H " lM.O.1 i t;' ';'f, TC-9 to

2. I m Steam

.T,*, *,'.;

Tc 4-h ,,Vake sen i Generator g j7r68?FC p3 c (73.$)' T -6 w.l.EJ I ,h,, M.,,I1; [iU Manual TC-5 TC-7 ses vake g,.. 3% I[d - ENC J 2.3 m 2 t21 TI.. liO f.3r'n, ..tyjd.:ay case 78R ' vake N'2 I?85LN5 I 5 :-.. wassam con e4m Note: TC's 1.8. and 9 are surface mounted on the pipe ad others are sensing infamal to the pepe. Figure 2-5 Waterhammer experiment (configuration #2).

I Point Beach NSAL-96-003 Evaluation Thermal / Hydraulic Analysis - Pre-Pump Start ) f - Analysis determined water column velocity based on s' team generation rate. - The Froude number is used to evaluate water column i velocity: U, l Fr= VgD - Based on experimental results,if the Froude number j approaches or exceeds unity, the horizontal pipe will run filled with water and significant condensation induced water hammer would not occur l = _ _ - u-i

Point Beach NSAL-96-003 Evaluation l Thermal / Hydraulic Analysis - Pre-Pump Start - Results for DBA show that: Amount of boiling is controlled by flow capacity of discharge lines CFC pressurizes to maximum of 43.7 psia >> Flow velocity is 4.2 ft/sec at 14 secs and 3.6 ft ft/sec at 30 i secs Froude number ranges from 0.91 to 0.78 which meets criteria a na

==i.* l

~\\ Point Beach NSAL-96-003 Evaluation Hydraulic Analysis - Post Pump Start - Model consists of three pumps, four coolers, corresponding valves, fittings and piping - Model based on actual plant configuration for elevations, number of fittings, and pipe lengths l - Pressure drop across cooler based on results of PBNP static flow model - All flow is assumed to go to upper coolers - Discharge lines are assumed to be drained to the throttle ) i valves outside containment l t =4" "

~ Point Beach NSAL-96-003 Evaluation l Hydraulic Analysis - Post Pump Start - Return lines will be filled with water in less than 20 seconds from start of first pump l - Re-fill rate in 8" return line will be less than 5 ft /sec 3 i r- - - _ a = l r

.t Point Beach NSAL-96-003 Evaluation . Piping Hydraulic Transient Analysis, Inputs and { Assumptions j i - Detailed force-time history developed for the model based on calculated re-fill rate: i >> Modelincludes return piping from CFC to throttle valve 3 A volumetric flow rate of 0.75 ft /see was input at each of the 21/2" CFC supply lines j >> It is postulated that shock will occur when water column j impacts the throttle valve t - SSE analysis performed with enveloped containment response spectra with ASME Code Case N-411 damping j

} i EVALUATION OF POSSIBLE i WATER HAMMER LOADS IN THE I POINT BEACH SERVICE WATER SYSTEM FOR DBA CONDITIONS i a Robert E. Henry Fauske & Associates, Inc. 4 Presented to i The International Joint Power i Generation Conference and Exposition i October 16, 1996 i l 1 \\MVos1016W.A

i I i i CONDENSATION INDUCED WATERHAMMER l This process is governed by the potential l for generating a large interfacial area between highly subcooled water and steam. l For a large interfacial area to exist, the l conditions must be consistent with a j stratified flow pattern in the pipe. i In the Point Beach cooling circuits, the j = piping is generally 8" diameter and the i nominal water velocity is 5 to 7 ft/sec. l i i } l f f i i HHWR101HA A I

l Q g

1. ~~

/ 7 / s \\ s / \\ 5 / \\ ( 1HX-15A } - 85 1 s c I 80 1 I } g g g t - 75 12-te ' i24 7 l l \\ i - 70 k 3 s I _l - 65 El 1 \\ El l il 5 - 55 g B' 0' g \\ -50 I 5 I I j C ~# I 15A R -40 I hl I 1 -15C l - 35 I Re u liDR. t 1 34-r ? - 30 I 1 X hJQ l I I - 25 3 7 pro 1 ( { l -20 1HX-15Cl ,_g gx.158 k 1 t ~' i From Unit 2j e 3 1HX-15D CFCS -10 I Unit 1 g 1HX-15A I SNib. / Unit 1 / SW ** 1 -5 g l -o 1 I To CW system To CW System em cwe s* l Figure 4 i t t n l

1 i DURING THE VOIDING TRANSIENT i i. Consider the heat transfer in a fan cooler to be 10 MW (34 MBTU/hr). i j The resulting steam velocity in the 8" pipe j is about 416 ft/sec (127 m/sec) which is an i order of magnitude greater than the j entrainment velocity (56.8 ft/sec /17.3 l m/sec) 1 4 l _ 3.7 gga (pr - P ) I est ~ /P i l Thus, unless the energy transfer is very low l or condensation on the heat sinks is very i high, there will be no stratification during l the voiding transient. 4 l i ] ~ 1,WYO\\l01696.A 4 I

l VOIDING IN THE FAN COOLER l i l Sustained steam generation in the fan cooler requires the presence of water. i 1 l If the voiding rate is too high, the steam l would " flood" the water in the cooling coils l and dynamically remove the water. This l would terminate the voiding leaving only column separation and rejoining. i i The maximum energy transfer that could occur in the fan cooler coils with water remaining to support continued steaming is about 1 MW. Thus, high voiding rates can not be sustained. LtHVO\\loteW.A

2-9 i i T,- 1 Water Initially a Discharged from the i i Fan Cooler Coils l "f' f Water Drainage Down the Discharge Piping: ] Because the Refill Velocity Exceeds the Bubble Rise Velocity, the Vertical Segrnent Essentially Runs Full of Water. i ) n.- . w i Water Drainage 4 Reaches the Horizontal 4 Piping: Horizontal Pipe i Does Run Full. Likely has a Two-phase Mixture at the Leading Edge. 2 - -= 1 4 .p. ;., j b Water Fills the Horizontal Pipe and Begins l . to Fill Vertical Piping. Likely has l 1 a Two-phase Mixture at the Leading Edge. l %,........-...-....9 RH96$031 tem 9-546 l Figure 2-3 Two-phase flow patterns during refill of the discharge piping, I '#APW3.2 1

__ 7 d 1 i 1 i TWO-PHASE FLOW i PATIERN CONSIDERATIONS For a Horizontal Pipe to Run Full i i l Froude No. (Fr) > 0.5 = U/U

• l U * = /gD i

If D = 8 in. (0.2 m), U* = 4.6 ft/sec (1.4 4 j m/sec). For a Vertical Pipe to Run Full i l U > U,,g, 1 .4 gD <'p' - p*q v2 i ) U,,g, = 0.345 = 0345 /gD Pt i i U,,g, = 0345 U

• i For D = 8 in., U, rise = 1.6 ft/sec (0.5 m/sec).

B 1 J f L\\MvG101696A

i 4 i i I TWO-PHASE FLOW l PATfERN CONSIDERATIONS l (Continued)

Conclusion:

For the conditions of interest, l the Froude Number greater j than unity a stratified flow l pattern will not exist in the l horizontal piping segments during the refill transient. 1 ufYelCleMA

i l 5 l 1 4 4 1 4 4 Q*p i TC 1 3.4 m H11.0 ft T-r- V, Y Solenoid i Operated ~ BaB Valves a 1 .-TC 2 4 s j 5.9 m Evacuated l N Receiver 2 i Vessel l --TC 3 Manual s.uvase e , y ~70 psig h P - -28.6 "Hg l Uo orifice i H TC 8 i .W w ,-z<nk $'E IEE TC-9 Gate 2.1 m i Steam -nWHj! Tc.4 - h 8jh p i Y l Generator 1Dd8 *F.! d

G:.20.2C..);

Tc-s p8 i .? ;7,':,;;;r; Manual C-5 Tc-7 g { .

• w 4 -- n

'd.' 5 J .".O.,;*.:0,, ;O...r Bau Vake g

- - f 4,

-250'rFE 1EE:,1i:-

2.3 m_ n. Gate 7.6 ft (i.121 ?C) u: . J.:.'::! !ir. art: :::; vaNo .m;;r..~,.; : -

• * '
• 9 t s' eweempie i.

b WB96A117.CoR 94 9G Note: TC's 1,8. and 9 are surface mounted on the pipe all others are sensing intamal to the pipe. e e Waterhammer experiment. N i

t WATER FPMMER EST #21 t e s t 5.ol - ( iii joiiij ,iisi igii jiii g.ii gi liisi; l i is i i i i o b Ten Completed 3: ~_ Column Separation [ p g, i ~~ e Normal Flow a_ Steam Added Ren11 t Initiated i~ O \\ t I If \\ NTest Apparatus l li .F iu -(i N [ [. y;p is Water Filled i,' li ii l il li l ii ,g i i i ii i i ! i ; ii i ii i i iii o ,e C 'C 20 30 40 50 50 70 8^ or 10C 11 0 12 0 " TIME" "CSEC:~ i

T o j l j l I i i I i I i ,6 i i i 9 4 <i i - o i : I : O 9 i 4 i t t t = O <s m r.n ~ ~ 00()o m (N HHHu = e c 4 _.-&. e o 6 g I to : t o' u to E ~ r w e C o,. I 1 1 M i ct ) lil i s O w 3 7 i 1 c J...........( t l 3 nm j .,, j O 1 k (N i-s _ _ _ _ _ _ _ _ ) o l 2, l, I I is i i o i i 0 0 <' 041 00t 04 0

0]

S3Idn000WH-!H1 9

6 AATER l-AMMER TEST tt21 o ia6 e 4 6 lt ii gi6 il4 i ! i l*i.ilt lI ij 6 6 et ii I I 6 L i I i 6 l4 o ,' tn v TC6" c "TC7" ---- "TC8"------ "FC9" O j y.= N, o i o if h;\\ .i \\ P, \\, ' r t. A Cc %' d \\\\ \\ y \\\\y ' s o 1 nn

/ 6

_ _h_ ~~y~~_,.,__ t i i l i l l O O 10 20 30 40 50 50 70 8C 90

CO 11 0 12 0

" T :ME" " C S E C)"

i i i 1 I CONCLUSIONS High steaming rates could not be suppmed l l for more than a few seconds. Without high l steaming rates, there is only a limited void j and the issue reduced to column rejoining, i f.e. determined by the refli! rate, i s Assuming high steaming rates gives l velocities much greater than the entrainment velocity. Thus, a stratified l condition could only exist due to the pipe j wall heat sink. i \\ Scaled experiments show only limited l waterhammer during the voiding period as l well as during the refill transient. l The pressurization when the refill water i reaches the throttle valve is very mild. I \\HWM01806 A t

4 NRC Generic Letter 96-06 Industry Meeting Hyatt Regency DFW j Dallas / Fort Worth Airport, Texas October 29,1996 Participants List Glenn Adams

• Biff Bradley Licensing Engineer Project Manager Wisconsin Electric Power Company Nuclear Energy Institute (414) 221 4691 (202) 739-8083 l

Mohammed F. Alvi Michael K. Brandon Structural Supervisor Senior Licensing Engineer Niagara Mohawk Power Corporation Entergy Operations, Inc. (315) 349-4121 (504) 381-4506 John Anciaux John J. Budans Senior Engineer Supervisor, Mechanical Engineering ) Wisconsin Electric Power Company Florida Power & Light Company (414) 221 2144 (561) 467 7117 Timotl. C 4 4 3reyebek Robert Campbell Advanad Technical Engineer Engineering Specialist Westinghouse Electric Corporation Tennessee Valley Authority (C2) 374 6246 (423) 751 8210 Asif H. Arastu Anwar Chaudhary Bechtel Power Corporation Design Engineer (415) 768 2247 Niagara Mohawk Power Corporation (315) 349-4052 Clint Ashley System Engineer Dan Collins Baltimore Gas and Electric Company System Manager (410) 495 4034 Baltimore Gas and Electric Company (414) 495 2370 Martin T. Babb Nuclere Specialist Randall M. Crane Southerr. Nuclear Operating Company Mechanical Engineering Supervisor (205) 992 7TP. Tennessee Valley Authority (205) 729 2561 Dennis W. Boyd Nuclear Safety and Licensing Specialist Charles E. Cronan Entergy Operations, Inc. Project Manager (501) 858-4616 Stone & Webster Engineering Corporation (617) 589 2357

m i Bill K. Dnk:: Steven Greco Engi :er Project Engineer, Mechanical Evaluations The Toledo Edison Company Wisconsin Electric Power Company (419) 321 8366 (414) 221 4387 i Jorge DelMazo Jack Hamm Engineering Supervisor Senior Mechanical Engineer Pacific Gas and Electric Company Florida Power & Light Company (415) 973-8326 (305) 246-6550 4 i Bryan A. Erler Garry Hammer Project Director Mechanical Engineer Sargent & Lundy U.S. Nuclear Regulatory Commission (312) 269 7132 (301) 415 279i Thomas C. Esselman Robert E. Henry President Senior Vice President j Altran Corporation Fauske and Associates,Inc. (617) 330-1130 (630) 323 8750 Charles K. Feist Jin-Shou Hseu Consulting Mechanical Engineer P ancipal Engineer TU Electric Wolf Creek Nuclear Operating Corporation (817) 897 8605 George Hubbard Stephen D. Fields Section Chief Supervisor, NSSS Design Team U.S. Nuclear Regulatory Commission Southern Nuclear Operating Company (301) 415 2870 (205) 992-6841 Paul Hubbard Michael G. Flynn Senior Engineer, Mechanical Evaluations Project Manager Wisconsin Electric Power Company Sargent & Lundy (414) 221 3668 (312) 269-6973 i Milton Huff Malcolm B. Garner Engineering Supervisor, Nuclear Design Senior Engineer Entergy Operations, Inc. Southern Nuclear Operating Company (501) 858 4326 (205) 992 6169 John Keelin Michael J. Gaucarz System Design Engineer Supervisor, Operations Analysis Southern California Edison Company ABB Combustion Engineering Nuclear Operations (714) 368-2310 (860) 285-4600 Frank W. Kennedy John Dennis Graves Senior Licensing Technologist Senior Engineer The Toledo Edison Company Entergy Operations, Inc. (419) 321 7909 (60l) 437 2341 Viswa S. Kumar Engineer, Plant Support The Detroit Edison Company (313) 586-1796 2

Jim Lee David W. Murphy Mechanical Evaluation Engineer Senior Engineer Wisconsin Electric Power Company Bechtel Power Corporation (414) 221 2313 (301) 417-3585 Marcia T. Lesniak Chalmer R. Myer Nuclear Licensing Administrator Manager, NSSS Design Team Commonwealth Edison Company Southern Nuclear Operating Company (630) 663 6484 (205) 992 6335 i Chris Ludlow An Nguyen Principal Engineer, Mechanical Engineering Senior Engineer, Stress and Metallurgy Baltimore Gas and Electric Company Southern Company Services. Inc. (410) 495 4107 (205) 992 7307 Daniel L. Magnarelli Kalyan K. Niyogi Senior Mechanical Engineer Director, Consulting Division Yankee Atomic Electri:: Company lioltec International, Inc. l (508) 568 2805 (609) 797 0900 L.B. (Tad) Marsh Chris Nolan Chief Plant Systems Branch Mechanical Engineer U.S. Nuclear Regulatory Commission Baltimore Gas and Electric Company j (30I) 415 2873 (410) 495 3654 t

• Paul D. Mason John M. Oddo Supervisor, Design Engineering Safety Assessment Connecticut Yankee Atomic Power Company Yankee Atomic Electric Company (860) 267 3131 (508) 568 2767 Eric W. May Bob Olech Supervisor, Mechanican Engineering Senior Engineer Virginia Power Southern California Edison Company j

(804) 273 2240 (714) 368-2309 John D. McCann William R. Peebles Licensing Section llead Manager, Power Systems Maine Yankee Atomic Power Company Sargent & Lundy (207) 798 4101 (312) 269-8600

• Mati Marilo Sid Powell Manager, Severe Accident Phenomenology Senior Nuclear Licensing Engineer Electric Power Research Institute Florida Power Corporation (415) 855 2104 (352) 563 4883 Dave Modeen Raubin Randels Director, Engineering Design Specialist Nuclear Energy Institute Commonwealth Edison Company (202) 739 8084 (630) 663-6615 3

Jon S. Ressler Vic Whaley Senior Nuclear Design Engineer Senior Licensing Project Manager Omaha Public Power Pistrict Tennessee Valley Authority (402) 533 6719 (423) 751 7009

• Paul H. Rothe Murr ay E. Williams Vice President Senior Mechanical Engii.eer Creare Inc.

Boston Edison Company (603) 643 4657 (508) 830-8275 Bill Rowlett Gil Williams Senior Staff Engineer Design Engineer Entergy Operations. Inc. South Carolina Electric & Gas Company (501) 858 4332 (803) 345-4159 Jim D. Seawright Roger Wilson Senior Licensing Engineer Senior Staff Engineer TU Electric Entergy Operations, Inc. (817) 897 0140 (501) 858-4305 William Selbe Steven D. Winter Project Engineer Senior Engineer Wolf Creek Nuclear Operating Corporation Consumers Power Company (316) 364 8831 (616) 764 2016 David E. Shafer Steven L. Wiser Supervising Engineer, Licensing Senior Staff Engineer Union Electric Company Virginia Power (314) 554-3104 (804) 273 2630 Avtar Singh George Woerner Manager, Plant Analysis and Safety Engineering Supervisor, Nuclear Design Electric Power Research Institute Entergy Operations, Inc. (415) 855-2384 (501) 856 4336 Jim Turkett Alan V. Wojchowski Licensing Engineer Superintendent, Safety Systems Engineering South Carolina Electric & Gas Company Northern States Power Company (803) 345 4047 (612) 295-1335 Bob Vasey Thomas E. Wroblewski Senior Engineer Senior Engineer, Modifications Engineering Group American Electric Power Service Corporation Wisconsin Electric Power Company (616) 697-5086 (414) 221 2802 Vaughn Wagoner Dale R. Wuokko Chief Mechanical Engineer Supervisor, llegulatory Affairs Carolina Power & Light Company The Toledo Edison Company (919) 546 7959 (410) 321 7120 Beth A.Wetzel Project Manager U.S. Nuclear Regulatory Commission Fpeakers (301) 415-1355 = 4

NRC Generic letter 96-06 Industry Meeting October 29,1996 8:00 a.m. - 5:00 p.m. Hyatt Regency DFW Dallas /Ft. Worth Airport, Texas A G E N D'A. -l l Tab 1 NRC Generic Letter 96-06 8:00 a.m. l Welcome & Introduction Biff Bradley, NEI Tab 2 l NRC Perspectives L.B. (Tad) hfarsh, NRR-PSB 8:15 a.m. Tab 3 George Hubbard, NRR PSB j Beth Wetzel, NRR-PD3 Garry Hammer, NRR-AfEB l 9:00 a.m. Questions & Answers for NRC ALL Break 10:15 a.m. 10.30 a.m. l Utility Actions and Lessons Learned Paul Afason, CY Tab 4 l Paul Rothe, Creare, Inc. l 11:30 a.m. Jorge del Afa:o, PG&E Tab 5 Lunch (provided) 12:00 p.nt I Utility Actions and Lessons Learned Steven Greco, WEPC 1:00 p.m. Tab 6 (Continued) hiichael Flynn, S&L Robert Henry, Fauske l 2:00 p.m. Vaughn Wagoner, CP&L Tab 7 Break 2:30 p.m. EPRI Activities to Support Industry Autor Singh, EPRI 2:45 p.m. Tab 8 Response to GL 96-06 Alati Aferilo, EPRI Discussion: Overpressurization of Biff Bradley, ALL 3:30 p.m. Isolated Piping Sections General Discussion ALL 4:00 p.m. l Tab 9 Meeting Participants List Adjourn 5:00 p.nt k}}