ML20202E643

Forwards Update to AP600 Final Ser,Chapter 20,GIs Addressing Issues for Which Srxb Has Primary Responsibility.Issues That Remain Open includes,A-31, RHR Shutdown Requirements, A-9, ATWS & A-105, Isloca

ML20202E643
Person / Time
Site:	05200003
Issue date:	01/09/1998
From:	Collins T NRC (Affiliation Not Assigned)
To:	Quay T NRC (Affiliation Not Assigned)
References
REF-GTECI-105, REF-GTECI-A-09, REF-GTECI-A-31, REF-GTECI-DC, REF-GTECI-NI, REF-GTECI-SY, TASK-105, TASK-A-09, TASK-A-31, TASK-A-9, TASK-OR NUDOCS 9802180207
Download: ML20202E643 (30)

Text

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January 9, 1998 s

O-MEMORANDUM TO: Theodore R. Quay, Director f2-3 Standardization Project Directorate Division of Reactor Program Management FROM: Timothy E. Collins, Chief / original aigned by/

Reactor Systems Branch Division of Systems Safety and Analysis EUBJECT: _ SRXB INPUT TO AP600 FSER, CHAPTER 20, GENERIC ISSUES Enclosed is our_ update to AP600 FSER, Chapter 20, Generic issues, addressing those

Issues for which SRXB has primary responsibility. Most issues are resolved based on .

recent responses from Westinghouse to staff RAls. However, some issues are remained

,. npen as follow: A-31, Residual Heat Removal Shutdown Requirements, A-9, Anticipated Transient Without Scram (AWTS), and issue 105, Interfacing System LOCA (ISLOCA).

If you have additional questions, please contact David Diec at 415-2834

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.FROM: Timothy E. Collins, Chief- '

Reactor Systems Branch j Division.of Systems Safety and , Analysis

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SUBJECT:

SRXB INPUT TO AP600--FSER, C PTER 20, GENERIC-ISSUES-Enclosed is our up ate to AP600 FSER, phapter 20, Generic Issues, addressing those issues for whibh SRXB has primary: responsibility. While most of the issues are resolved b'ased on recent iesponses from Westinghouse to staff RAls, some issues remain ope'tg.. Upon co etion of_the review of Open Issues, we will provide a suppleme'nt to Cha fer 20.

If you have additional q\ s ti 's, please contact David Diec at 415-2834

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lasue A-31: Residual Heat Removal Shutdown Requirements As discussed in NUREG-0933, lasue A 31 addressed the ability to transfer heat from the reactor to the environment after shutdown, which is an important safety function, it was resolved in 1978 with the issuance of SRP Section 5.4.7 " Residual Heat Removal (RHR)

System."

--The safe shutdown of a nuclear power plant following an accident not related to a LOCAL  !

has typically been interpreted as achieving " hot standby" condition. The NRC has placed considerable emphasis on the hot-standby condition of a power plant in the event of an accident or other abnormal occurrence and, s!mliarly, on leng-term cooling, which is

- typically achieved by the recidual heat removal (RHR) system. The RHR system starts to operate when the reactor coolant pressure and temperature are substantially lower than i

. their hot-standby-condition values. Even though it may generally be considered safe to maintain a reactor in hot-standby condition for a long time, experience shows that certain

. events have occurred that required eventual cooldown or long-term cooling until the RCS  :

le cold enough for personnel to inspect the problem and repair it.

In SSAP Section 1.9.4.2.2, Westinghouse stated that the AP600 design includes passive safety-related decay heat removal systems that establish and maintain the plant in a safe-shutdown condition following design + - is events and it is not . 4cessary that these -

! . rassive system achieve cold shutdown as defined in RG 1.139.-

- The passive core cooling system is design to maintain plant safe-shutdown conditions indefinitely. Cold shutdown condition is necessary only to gain access to the reactor coolant system for inspection, maintenance, or repair. For the AP600 design, cold shutdown conditions can be achieved using highly reliable, but non safety-related sys-tems, which have similar redundancy as current generation safety-related systems and ce supplied with ac power from either onsite or offsite sources. Passive core cooling capability is ditoussed in Section S.3 of the SSAR.

Westinghouse states that the passive residual heat removal system can achieve hot

- standby conditions immediately and can reduce the reactor coolant temperature to '

215.6 *C (420 'F) within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The reactor pressure is controlled and can be reduced

' to 1.72 MPa (250 ps!g). The passive RHR system also provides a closed cooling system to maintain long-term cooling. Therefore, the AP600 complies with GDC 34 by using a more -

reliable and simplified system for bo'n hot standby and long-term cooling modes, and it is

- not necessary that these passive syt.tems achieve cold shutdown as defined by RG 1.139.

GDC 34 requires a residual heat removal system to be provided with suitable redundancy .

In components and features to assure that, with or without onsite or offsite power, it can accomplish its safety functions so that the specified acceptable fuel design ilmits and the design conditions of the reactor coolant pressure boundary are not exceeded. No definition is specified as the safe-shutdown condition for which the RHR syrnm should accomplish.- EPRIin the Utility Requirements Document for F'assive ALWRb proposed that the safe-shutdown condition be defined as 215.6 *C (420 'F) for the passive ALWR

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2 designs. The staff has concluded that cold shutdown is not the only safe stable shutdown condition which can maintain the fuel and reactor pressure boundary within acceptable limits, in SECY-94 084, Section C, " Safe Shutdown Requirements," the staff recom-mended, and the Commission approved, that the EPRl's proposed 215.6 'C (420 *F) criteria or below, rather than the cold shutdown condition required by RG 1.139, be accepted as a safe stable condition, which the passive RHR system must be capable of achieving and maintaining following non-LOCA events.- This acceptance is predicated on an acceptable -

passive safety systern performance and an acceptable resolution of the issue of regulatory  ;

treatment of non safety systerns (RTNSS). The SECY paper also states that the passive safety system capabilities can be demonstrated by appropriate evaluations during detailed design analyses, including:

(1) A safety analysis to demonstrate that the passive systems can bring the plant to a c safe stable condition and maintain this condition, that no transients will result in the specified acceptable fuel design limits and pressure boundary design limit being violated, that no high4nergy piping failure being initiated from this condition will result in violation of 10 CFR 50.46 criteria; and (2) A probabilistic reliability analysis, including events initiated from the safe-shutdown conditions, to ensure conformance with the safety goal guidelines. The PRA would -

also determine the reliability / availability missions of risk significant systems and components as a part of the effort for regulatory treatment of non safety systems.

The resolution of issue A-31 for the AP600 design remains open as the staff is evaluating both the passive system parformance capability through testing and safety analyses, and the proper resolution of the RTNSS issue.

. Issue A-9: Anticipated Transient Without Scram As discussed in NUREG-0933, issue A-9, addressed the issue of ensuring that the reactor ,

can attain safe shutdown after incurring an anticipated transient with a failure of the reactor trip system (RTS). An anticipated transient without scram (ATWS) is an expected operational yccurrence (such as loss of feedwater, loss of condenser vacuum, or loss of offsite power (LOOP) to the reactor) that is accompanied by a failure of the RTS to shut

- down the reactor.

The acceptance criterion for the resolution of issue A-9 are as follow:

_ e Compliance with the mitigation requirement of 10 CFR 50.62(c)(1) that plant equip-ment must automatically initiate emergency feedwater (EFW) and turbine trip under conditions indicative of an ATWS. This equipment must function reliably and must be diverse and independent from the RTS.

e Compliance with the prevention requirement of 10 CFR 50.62 (c)(2) that the plant must have a scram system that is diverse and independent from the existing RTS.

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3 in SSAR Section 1.9.4.2.2, Westinghouse stated that the AP600 design complies with the requirements in 10 CFR 50.62, " Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants," and a discussion of the design features to address the probability of an ATWS is included in

Sections 1.9.5 and 7.7 of the SSAR.

Westinghouse indicated that the AP600 design complies with the requirements of 10 CFR 50.62 with a diverse actuation system that includes the AMSAC (ATWS mitigation system -

actuation circuitry) protection features mandated by 10 CFR 50.62 by tripping the turbine

. and diversely actuating selected engineered safeguards functions. However, the AP600 design dos., not automatically initiate auxiliary FW, but instead includes a atomatic -

initiation of the PRHR cooling system. _The AP600 design does not, therefore comply with 10 CFR 50.62.

There are other AP600 design features aimed at minimizing the probability of ATWS I

occurrence and mitigating the consequences, as discussed in Section 1.9.5 of the SSAR.

For the AP600 design with passive core cooling systems, the staff requires that an ATWS analysis be performed to demonstrate that its ATWS response is consistent with that -

considered by the staff in its formulation of the 10 CFR 50.62 design requirements for current plant designs. :The applicant has provided (response to RAI 440.26) the analysis of a complete loss of normal feedwater without reactor trip, using the LOFTRAN code. The -

- staff has requested additionalinformation concerning the AP600 ATWS analysis, which the applicant has agreed to provide. This is an open issue. The use of the PRHR system in-lieu of automatic AFW is also an open item.

Therefore, lasue A-9 is not resolved for the AP600 design.

Issue A-17: Systems Interactions in Nuclear Power Plants

' As discussed in NUREG-0933, issue A-17 addressed concerns regarding adverse sp xms :

Interactions (ASIS) in nuclear power plants.~ Depending on how they propagate, Asis can be classified as functionally coupled, spatially coupled, and induced-human-intervention coupled. As discussed in NUREG-1229, " Regulatory Analys!s for Resolution of USl A-17,"

datad August 1989, and GL 89-18, " Resolution of Unresolved Safety issue A-17, Systems Interactions in Nuclear Power Plants," dated September 6,1989, lasue A-17 concerns Asis caused by water intrusion, internal flooding, seismic events, and pipe ruptures.

A nuclear power plant comprises numerous structures, systems, and components (SSCs) that are designed, analyzed, and constructed using many different engineering disciplines.

The 6 gree of functional and physical integration of these SSCs into any single power plant may vary considerably. Concerns have been raised about the adequacy of this  ;

functional and physicalintegration and coordination process. The issue A-17 program was '

Initiated to integrate the areas of systems interactions and consider viable alternatives for regulatory requirements to ensure that the Asis have been or will be minimized in operat-Ing plants and new plants. Within the framework of the program, the staff requested, as i

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stated in NUREG-0933, that plant de' signers consider the operating experience discussed in GL 89-18 and use the probabilistic risk assessment (PRA) requ! red for future plants to identify the vulnerability and reduce Asis.

This issue identified the need to investigate the potential that unrecognized subtle

= dependencies, or systems interactions, among structures, systems, and components (SSCs) in a plant could lead to satety significant svents. in NUREG-1174, intersystem dependencies are categorized based on the way they propagate into functionally-coupled, spatially-coupled, and induced human-inteNontion coupled systems interactions. The.

occurrence of an actual adverse systems interaction (ASI) or the e.tistence of a potential-ASI, as well as the potential overall safety impact, is a function of an individual plant's

design and operational features. For AP600 with new or differently configured passive and l active systems, a systematic search for Asis is necessary.

Westinghouse submitted WCAP 14477, Revision 0,,"The AP600 Adverse System Interac-l tion Evaluation Report," dated February 19fr6 for staff ruview and approve. Tne purpose

- of the report was to identify possible adverse interactions among safety-related systems and between safety-related and non-safety-related system 3, and to evaluate the potential ,

consequences of such interactions. The staff reviewed WCAP 14477 report and provided .

" Westinghouse with comments and questions. Westinghouse subsequently addressed the staff's questions and comments and issued a revision to the WCAP-14477 report. The staff reviews this issue as part of the regulatory treatment of non-safety systems (RTNSS) and has documented its mview in Chapter 22 of the FSER. '

The staff concludes that Westinghouse has adequately assessed possible adverse systems interactions and their potential consequences in WCAP-1447, revision- 1. In L addition, the staff has conducted confirmatory testing involving potential systems interac-tions, and has performed analyses of selected accident scenarios in which nonsafety and/or safety systems could interact. Both the confirmatory tests and analyses showed that potential systems interactions did not have significant adverse effects on overall .

safety performance. Additionally, no additional unanticipated adverst systems interac-

- tions were observed. Ols 20.2-5 and 20.2-6 are closed. This issue is considered closed.

Issue A-26: Reactor Vessel Pressure Transient Protection Since 1972, there have been many reported pressure transients which have exceeded the pressure-temperature limits specified in technical specifications (TS) for PWRs. The majority of these events occurred at relatively low reactor vessel temperatures at which the material has less toughness and le more susceptible to failure through brittle fracture.

This is issue A-26 in NUREG-0933 which was resolved with the issuance of SRP Sec-tion 5.2.2," Overpressure Protection." Applicants for cps and operating licenses were requested to design an overpressure protection system for light-water reactors (LWRs) following the guidance provided in SRP Section 5.2.2.

jn its May 28,1993, letter, Westinghouse stated that the APC00 dedgn conforms to the criteria in Branch Technical Position (BTP) RSB 5-2, "Overpressurizanoc. Protection of Pressurized Water Reactors While Operating at Low Temperatures," of SRP Section 5.2.2.

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The pressurizer is sized to accommodate most pressure transients and over pressure protection for the RCS is provided by either the pressurizer safety valves or the normal residual heat removal relief valves, as described in Section 5.2.2 of tae SSAR.

The staff concludes that AP600 design satisfies the BTP ,RSB 5-2 requirements and therefore, considers this issue closed.

Issue B-22: LWR Fuel Westinghouse identified in Table 1.9-2, of its May 28,1993, letter that it considered issue B-22 relevant to the AP600 design; however, this issue is not required for the AP600 design to meet 52.47(a)(1)(ii) or(iv).

As discussed in NUREG-0933, Issue B-22 addressed the staff concerns that Individual reactor fuel rods sometimes failed during normal operations and many fuel rods are expected to fall during severe core accidents. Failure of fuel rods results in radioactive releases within a plant and is a potential source of release to the public. The resolution of

- this issue was to ensure that these fuel failures did not result in unacceptable releases to

- the public. Several problem & were identified in the staff effort to improve the predictability L of fuel performance and these were addressed in the revision to SRP Section 4.2, " Fuel System Design,"in 1981. The staff concluded that the then existing requirements on fuel were adequate to ensure continued low fuel defect rates and additior.al requirements would not significantly decrease the number of fuel defects. This issue was then dropped from '

further consideration.

Westinghouse stated the AP600 reactor core design complies with SRP Section 4.2 and the discussion on the fuel system design is in Section 4.2 of the SSAR.

l The staff has completed its review of the VANTAGE-5H fuel for the AP600 design.' The details of fuel design and acceptance criteria are discussed in Section 4.2 of the final-c safety analysis report. The staff concludes that Westinghouse has satisfactorily resolved all quostions raised during the staff review of the issue, and therefore, the staff considers this issue resolved. The Open item 4.2.8-1 is closed, issue C-4: Statistical Methods for ECCS Analysis As discussed in NUREG-0933, Issue C-4 addressed the statistical methods used for perfor-mance evaluation of the ECCS during a LOCA. In accordance with the requirements of

- 10 CFR 50.46 as amended on September 16,1988, the NRC requires that the LOCA analyses for license applications use either the 10 CFR Part 50 (Appendix K) evaluation ,

models or the statistical (realistic) models, including the uncertainty of calculation in the '

adverse direction.- The realistic models must be supported by applicable experimental data. Uncertaintlas in the realistic models and input must be identified and assessed so that uncertainty in the calculated results can be estEnated.

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In SSAR Section 1.9.4.2.2, Westinghouse stated the AP600 methodology appliad for LOCA analysis is discussed in SSAR Chapter 15.

- Appendix K of 10 CFR Part 50 specifies the requirements for LWR ECCS analyt is, which call for specific conservatism to be applied to certain moc%Is and correlations used in the analysis to account for data uncertainties at the time Apendix K was written. US; C-4

' addresses NRC development of a statistical assessment of the uncertain level of the peak cladding temperature limit. - in 1988,10 CFR 50.46, " Acceptance Criteria for ECCS for Light -

Water Nuclear Power Reactors," was revised to allow the realistic ECCS evaluation model, .

In addition to the evaluation model conforming to the Appendix K requirements. This BE evaluation model will use analytical technique realistically describing the behavior of the reactor system during a LOCA, with comparisons to applicable experimental data. -The realistic evaluation model must identify and account for uncertainties in the analysis method and inputs so that when the calculated ECCS cooling performance is compared to the acceptance criteria, there is a high level of probability that the criteria would not be ext:eeded.

As described in SSAR t.hapter 15, computer codes WCOBRAITRAC and NOTRUMP, respectively, are used for the large- and small-break LOCA analyses. WCOBRAITRAC is a realistic code, and the uncertainties will be included in the analysis. NOTRUMP is a code using the Appendix K requirements.

Issue C-4 is closed.

- Issue C-5: Decay Heat Update As discussed in NUREG-0933, issue C-5, addressed the specific decay heat models for the LOCA analysis models. In ecordance with the requirements of 10 CFR 50.46 as amended on September 16,1688, the LOCA analyses for license applications should use either the 10 CFR Part 50 (Appendix K) models, or the realistic models suppcdsd by applicable experimental data and including uncertainty of calculation in the aaverse directio' . When Appendix K models are used, the docty heat generation function should be based on ANS 5.0," Decay Energy I4elease Rates Following Shutdown of Uranium-Fueled Thermal Reac-tors," plus a 20-percent uncertainty factor. When realistic models are used, the decay heat function in ANS 5.1, " Decay Heat Power in Light Water Reactors," is acceptable for licensing applications, in SSAR Section 1.9.4.2.2, Westinghouse stated that the large-break LOCA analyses for the AP600 design, discussed in Section 15.6.5 of the SSAR, used the decay heat model identified in the 1979 ANSI 5.1 standard.

This issue involved following the work of resear::h groups in determining best-estimate decay heat data and assot.inted uncertainties for use in LOCA calculations.

Appendix K of 10 CFR Part 50 requires the use of 1971 ANS Standard, " Decay Energy ,

Release Ratcs Following Shutdown of Uranium-Fueled Thermal Reactors," times 1.2 be used for the heat generation rates from the radioactive decay of fission products in the

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-. ECCS calculation. _The staff aas determined that the 1979 ANSI 5.1 is technically accept-able and has allowed this use in the realistic evaluation model. For the AP600 application, the 1971 ANS decay heat model and the 1979 ANSI decay heat model are used in NOTRUMP and WCOBRA/ TRAC, respectively, for small and large-break LOCAs. The staff has completed and documented its review of WCOBRAITRAC and NOTRUMP in Chapter 15 of the AP600 FSER.' The staff considers issue C-5 closed.  ;

-Issue C-6: LOCA Heat Sources

. As discussed in NUREG-0933, Issue CC addressed the issue identified in NUREG-0471 which involved staff evaluations of vendors' data and approaches for determining LOCA heat sources and the need for developing staff positions. - The contributors to LOCA heat sources, along with their associated uncertainties and the manner in which they are combined, have an impact on LOCA calculations. the staff informed the Commission in L SECY-83-472," Emergency Core Cooling System Analysis Methods," November 17,1983, i, that statistical combination of LOCA heat sources would be allowed to justify the relaxation v of non-required conservatism in emergency core cooling system (ECCS) evaluation v

- models.

in SSAR Section 1.9.4.2.2, Westinghouse stated that the discussion of LOCA heat sources for the AP600 design is inc!~ied in Section 15.6.5 of the SSAR.

The staff has completed anu ocumented its review of WCOBRAITRAC and NOTRUMP in Chapter 15 of the AP600 FSER. The staff considers issue C-6 closed.-

Issue 22: Inadvertent Boren [ . tion Events

As' discussed in NUREG 0933, issue 22 addressed the possibility of core criticality during .

cold shutdown conditions from inadvertent boron dilution events.- Although this issue was ru olved with no new requirements, the acceptance criterion is that plants shall minimize the consequences of such events by mee, ng SRP Section 15.4.6, " Chemical and Volume.

. Control System Malfunction that Results in i Decrease in Boron Concentration in the Reactor Coolant _ (PWR)."' Specifically, the_ plcnt shall respond in such a way that the . _

criteria regarding fuel damage ano system pressure are met, and the dilution transient is terminated before the shutdown margin is eliminated. If operator action is required to terminate the transient, redundant alarms must be in place and the following minimum time

_c intervals must be available between when an alarm announces an unplanned dilution and when shutdown margin is lost:.

e. during refueling (Mode 6) - 30 minutes e during all other operating modes - 15 minutes Section 15.4.6 of the SSAR provides a safety analysis which demonstrates that redundant alarms are available to enable operators to detect und terminate an inadvertent boron dilution event within the above required time intervals, before shutdown margin is lost.

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In addition to the events in this issue, the staff has identified the following two boron

, dilution scenarios where a deborated water slug may accumulate in the RCS and a restart ,

of the RCPs will cause this slug to pass through the core resulting in criticality or a power excursion: ._

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e. The first scenario occurs during a plant startup when the reactor is deborated as part of startup procedures. A loss of offsite power will result in tripping the RCPs and charging pump. The subsequent startup of the diesel generator will restart the charging pump and cause the accumulation of deborated water in the reactor lower plenum. The RCP restart with recovery of offsite power will cause this deborated water to pass through the core.

e The second scenario is related to transients or accidents, such as a small break LOCA with heat removal by reflux condensation natural circulation that may result in

an accumulation of deborated water in the RCS loop. This water will pass through the core with an inadvertent restart of the RCPs. <

The staff has completed and documented its review of inadvertent boron dilution issue in-Section 15.2.6.5.4 of the staff FSER. The staff considers issue-22 closed. The Open item

. 20.3-2 is closed.

Issue 105: Interfacing System LOCA (ISLOCA) at LWRs issue 105, l' NUREG 0933, was limited to pressure isolation valves (PlVs) in BWRs and was resolve 1 bv Aquirir g leak-testing of the check valves that isolate low-pressure systems that are connected at the RCS outside of containment. It is related to issue 96 which addressed PlVs between the RCS and RHR systems in PWRs. As stated in NUREG 0933, the staff issued Information Notice (IN) 92-36, intersystem LOCA Outside .

Containment," dated May 7,1992, on this subject. The individual plant examinations required by the stati on operating plants included analyses of these sequences. This issue was resolved without any new requirements for operating plants.

2 The staff position regarding intersystem LOCA protection, as stated in SECY 90-016,

" Evolutionary Light Water Reactor (LWR) Certification lasues and Their Relationship to Current Regulatory Requirement," as well as SECY 93-087, " Policy. Technical, and Licensing issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR)

Designs,"is that ALWR designs should reduce the possit9ity of a LOCA outside contain-ment by designing, to the extent practicable, all systems and subsystems connected to the RCS to an ultimate rupture strength (URS) at least equal to full RCS pressure. The phrase

- "to the extent practicable" is a recognition that all systems must ever:tually interface wit 5 atmosphere and that it would be difficult or prohibitively expensive to design certain large

< tanks and heat exchangers with the URS equal to full RCS pressure. The degree of isolation or a number of barriers is not a sufficient justification for using low-pressure components that can be practically designed to the full RCS URS criteria. Piping runs should be designad to meet the URS criteria, as should all associated flanges, connectors, and packings, including valve stem seals, pump seals, heat exchanger tubes, valve

9 bonnets, and RCS drain and vent lines. The desif 7er should attempt to reduce the level of pressure challenge to all systems and subsysten.4 connected to the RCS.

In Section 3.9.3.1 of this report, the staff dircusses its evaluation which establishes the minimum pressure for which low-pressure systems should be designed to ensure reasonable protection against burst failure, should the low-pressure system be subjected to full RCS pressure. The subsection within Section 3.9.3.1, "AP600 Design Criteria for ISLOCA," contains the design criteria proposed by Westinghouse for the low-pressure -

portion of the normal residual heat removal system (RNS). On the basis of this evaluation, the staff concludes that this criteria is acceptable to ensure that the low-pressure side of any applicable system has been designed to meet the full RCS URS criteria.

For all interfacing systems and components that do not meet the full RCS URS criteria, the -

uplicant must justify why it is not practicable to reduce the pressure challenge any further, and also provide compensating isolation capability. For example, applicants should demonstrate that for each interface the degree and quality of isolation or reduced severity of the potential pressure challenges are compensated and justified for the safety of the low-pressure interfacing systems or components. The adequacy of pressure relief-and the piping of relief back to primary containment are possible considerations. As identified in SECY-90-016, each of these interfacing systems that has not been designed to withstand full RCS pressure must also include the following protection measures: -(1) the capability for leak testing of the pressure isolation valves, (2) valve position indication that is available in the control room when isolation valve operators are de-energized, and (3) high-pressure alarp s to warn control room operators when rising RCS pressure approaches the design pressure of the attached low-pressure system and both isolation valves are not closed.

Section 1.9.5.1.7 of the SSAR discusses Westinghou 6e responses regarding compliance of -

the AP600 design with the staff position on ISLOCA. In WCAP-14425, " Evaluation of the AP600 Conformance to Inter-System Loss-of-Coolant Accident Acceptance Criteria," dated July 1995, Westinghouse performed a systematic evaluation of the design responses of various systems interfacing the RCS of the ISLOCA challenges. - The systematic evaluation process includes (1) a review of the AP600 piping and instrumemation diagrams to identify these primary interfacing systems or subsystems directly interfacing with the RCS, and the sucondary interfacing systems or subsystems interfacing with the primary interfacing systems, (2) identification of primary and secondary systems and subsystems having its URS less than the RCS pressure. For those systems or subsystems not meeting the criterion _of the URS greater than the RCS pressure, a design evaluation is made on whether it is inside containment, whether it meets the three criteria specified in SECY 90- ,

016, and or whether it includes other design features specific to them that prevent an ISLOCA to the extent practicable. The report also provides its reasons why it is not practical to design large, low-design pressure tanks a.id tank structures that are vented to the atmosphere to the high pressure criterion. Interfacing systems or subsystems that connect directly to an atmospheric tank are excluded from further ISLOCA consideration.

This is limited to the piping connected directly to the atmospheric tank, up to the first isolation valve other than a locked-upon, manual isolation valve. The staff evaluation of these systems follows.

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10 (a) Normal Residual Heat Removal System:

The portions of the RNS from the RCS to the containment isolation valves (CIVs) outside containment are designed to the RCS operating pressure, and the portions downstream of the CIVs and upstream of the discharge line CIVs are designed so that its URS is not less than the RCS operating pressure. The mechanical shaft seal of the RNS pump with a design pressure of 900 psig is the only portion of the RNS having the URS lower than the RCS pressure. Subsection 3.1.3.2 of WCAP-14425 discusses the difficuttles of designing the RNS pump seal to full RCS operating pressure. A fundamental problem is that any type of seal that can withstand the full RCS pressure willlikely have abnormally fast wear of the seal faces during normal plant operation at low seal pressure. This increased wear at normal plant operating conditions could well prevent the seal from maintaining the pressure boundary if ever exposed to the full RCS pressure. The use of high pressure seals will alsr> r=@e more frequent maintenance during normal operation. Therefore, it is impractical to design a seal that would maintain the RCS pressure boundary with no leakage, and also opercte satisfactorily at low-pressure conr tions. The AP600 RNS pump mechanical seal is desir,ned to minimize the amount oi leakage if exposed to full RCS pressure. An INEL study on the Davis Besse Nuclear Power Station decay heat removal

- pump seal, with a design pressure of 450 psig, found that the rotating seal wr( 9d maintain its structuralintegrity to pressure in excess of 2500 psi, and the mechanical seals could withstand a pressure of 1200 to 1250 psi without leaking. The AP600 RNS pump mechanical seal is similar to the Davis Besse DHR pumps, however, its design pressure is twice as high. The AP600 RNS pump also has a disaster bushing that limits the leakage from the pump to within the capabilities of the normal makeup system in case of catastrophic mechanical seal fai'are. The leakage can be controlled with the sealleakoff line routed to a floor drain and subsequently to the auxiliary building sump. This is more favorable than a seal specially designed for full RCS pressure at the expense of normal-condition reliability.

Subsection 5.4.7.2.2 of SSAR discusses the AP600 design features in the RNS specifically aimed at reducing the likelihood of an intersystem LOCA. On the suction side, there is a normally closed motor-operated isolation valve in the common suction line outside containment, and two normally closed motor-operated isolation valves in each parallel suction line inside the containment. There is also a relief valve with a set pressure of 563 psig connected to the RNS pump suction line inside containment. This valve is designed to provide low-temperature overpressure protection of the RCS and will reduce the risk of overpressusizing the RNS. On the discharge side, the common discharge line has a safety-related containment isolation check valve inside containment and a safety-related motor-operated isolation valve outside containment. The MOVs inside the containment are interlocked to prevent them from opening when the RCS rassure is above the RNS operating pressure of 450 psig. The power to these isolation valves is administratively blocked at the valve motor control centers to prevent inadvertent opening, in addition, the discharge header contains a relief valve, which discharges to the liquid radwaste system (WLS) effluent holdup tanks, to prevent overpressure in the RNS pump discharge line that could occur if the three check valves and the motor-operated CIV leaked back to the low pressure portions of the RNS.

s - - _ _ _ _ _ - _

11 Also, the RNS design includes an instrumentation channel that ;ndicates pressure in each 1 RNS pump suction line, and a high pressure alarm is provided in the main control room to alert the operator to a condition of rising RCS pressure that could eventually exceed the RNS design pressure. The motor-operated pressure isolation valves also have remote position indications in the main con +rol room. However, the RNS design does not include the capability to conduct leak testing of the PiVs as required by SECY 90 016 for those systems not designing with the URS capable of withstanding the RCS pressure. AP600 technical spcMfication LCO 3.4.16 requires the integrity of each RCS PlV be maintained with surveillance requirements of the PIV to verify its operability specified in SR 3.4.16.1 and 3.4.16.2. However, the TS Basis excludes the RNS PlVs from the TS 3.4.16, which is not consistent with SECY-90-016. The staff will_ require the RNS PlVs be included in TS 3.4.16 for periodic leak testing. This is an Open item.

(b) Chemical and Volume Control System - Makeup systems:

Subsection g.3.6 of the SSAR provides a detailed description of the design, functions, and operations of the chemical and volume control system (CVS). The purification flow path of the CVS is a high-pressure closed-loop design which is entirely within the containment.

The patential contributors to an ISLOCA are the portions of the CVS located outside the containment, i.e., the letdown line to the liquid radwaste system, and the makeup system.

The CVS makeup pumps operate intermittently to make up for RCS leakage. The pumps start and stop automatically when the pressurizer level reaches the bottom and the top of the normal level band, respectively. The makeup pumps take suction from either the boric acid tank, or the domineralized water storage tank (DWST), and inject into the CVS -

purification loop return stream. The makeup pumps can also take suction from the waste

- holdup tanks or the spent fuel pool. The makeup line from the makeup pump discharge to '

the RCS has a design pressure greater than or equal to the RCS design pressure.

However, the pump suction line piping and associated components have a design pressure of 150 psig with the URS less than the RCS operating pressure.

Subsection 3.3.3 of the WCAP-14425 contends that it is not practicable to design the low-pressure portions of the makeup suction piping to higher design pressure. It is not practicable to have a high design pressure for large tanks such as the boric acid tank,

' which are vented to the atmosphere, as well as the piping directly connected to these atmospheric tanis up to the first isolation valve. The suction lines each ccntain a check valva that separates the suction piping from a large atmospheric tank. These check valves are designed to open on low differantial pressure, and have a high tendency to leak. If the two discharge line check valves are assumed to leak, then it is also reasonable to assume that the suction line check valves willleak. The efore, designing the suction piping to a

- higher pressure will only increase the likelihood that the RCS leakeage will flow to one of the atmospheric tanks. The suction lines contain relief valves that protect the low- .

pressure portions of the piping from overpressure in the event of leaking check valves in the discharge line or thermal expansion in case of a loss of mlniflow cooling. .'he relief valves direct any leakage from the discharge line check valves to the WLS effluent holdup tanks (EHTs), which is designed to handle radioactive fluids, and its level is monitored by remote 'nstrumentation.

12 The pcssage of the hig'. >. essure reactor coolant to the CVS makeup suction is possible only when the makeup pumps are not running, or as a result of failures or leakage of multiple check valves on the makeup pump discharge side. There is a high-pressure alarm in the pump suction line to alert the operator of overpressurization. In the event of a suction-slde overpressurization, the makeup pumps can be operated to terminate overpressurizing the suction piping, if the makeup pumps do not .. art, the makeup line containment isolation valves would automatically close to terminate the ISLOCA. In addition, the purification irop inlet isolation valves will also be closed on a safeguards actuation signal. These multiple, safety-related isolation valves mitigate an ISLOCA in the makeup suction line. The makeup line CIVs have the capability for leak-testing, and are provided with valve position indication in the control room at all times. The staff finds that protection measures meet the intent of SECY-90-016 ISLOCA position with the following exception of the CIV leak testing. The makeup line CIVs are also the pressure isolation valves. They should be subjected to the PlV leak test requirements, and be included in TS 3.4.16 for PlV LCO and surveillance requirements.

WCAP-14425 does not provide the relieving capacity and setpressure of the makeup system relief valves, as well as the analysis to demonstrate the acceptability of these values. The staff will require that the relieving capacity and setpressure of the relief valves in the makeup system along with their bases be documented. This is an Open item.

(c) CVS Letdown - Liquid Radwaste System:

The CVS letdown line connects to the high-pressure purification loop inside containment.

Immediately downstream of this connection is a high-pressure, multi-stage letdown orifice, which reduce pressure in the letdown line from the RCS operating pressure to below the design pressure of the low-pressure portion of the letdown line. Around the letdown orifice is a bypass line containing a locked-closed manualisolation valve that is opened only at chutdown when the RCS is depressurized to provide sufficient letdown flow when required.

The letdown line is then equipped with two safety-related, normally-closed, fail-closed CIVs while penetrating containment to the liquid radwaste (WLS) degasifier package and EHTs.

The letdown line down to and including the outboard CIV has a design pressure of 2485 psig. Downstream of the outboard CIV, the WLS letdown line has a design pressure of 150 psig, and therefore does not meet the RCS URS criteria.

Subsection 3.2.3 of WCAP-14425 contends that it is not practicable to design the low pressure portions of the letdown line to a higher design pressure. The WLS EHTs are large atmospheric tanks, and are therefore not practicable for higher design pressure. Nor are the letdown line, which is routed to the degasifier package or the EHTs, and the degasifier package, which discharges directly to the WLS EHTs. The CVS letdown system has the following features to meet the ISLOCA criteria: (1) the pressure drop across the CVS letdown orifice protects the WLS from overpressurization during letdown operations by reducing the pressure in the WL6, (2) in case of an inadvertent valve closure in the WLS during letdown, a relief valve, which discharges directly to the EHT, is provided that would protect the WLS from overpressurization, (3) due to the letdown orifice, a break in the WLS during letdown from the CVS would result in an RCS leak that is within the capability of the normal makeup system,(4)if an ISLOCA should occur, it would be terminated by automatic

13 isolation of the two purification loop isolation valves and two letdown isolation valves on low pressurizer level or a safeguards actuation signal, and (5) the letdown line CIVs have the capability for leak testing and have valve position indication in the control room at all time, and (6) the WLS degasifier column contains a high pressure alarm that would warn the control room operators that the WLS pressure was approaching the design pressure.

The staff finds these arguments acceptable except for the following: The CIVs are also the pressure isolation valves, and should be subjected to the PlV leak testing requirements, instead of leak testing for CIVs. These valves should be included in the AP600 technical specifications LCO 3.4.16 and associated surveillance requirements. This is an Open item.

(d) Primary Sampling System:

The primary sampling system (PSS) collects representative semples of fluids from the RCS and associated auxiliary system process streams, and the containment atmosphere for analysis by the plant operating staff. Section 3.4 of WCAP-14425 provides an ISLOCA evaluation of the PSS. The PSS pipings are 3/8-inch small pipes. The whole PSS is designed to full RCS pressure and temperature, with the exception of the following low pressure portions: eductor water storage tank (EWST) and its drainage and level indication lines, eductor supply pump seal, and demineralized water supply line. These portions have design pressures with an URS below the RCS operating pressure. The applicant contends that it is not practical to design the low pressure portion of the PSS to a higher design pressure because they are at atmospheric pressure and connect to the low pressure demineralited water system (DWS). Designing the EWST to high pressure to meet ISLOCA criteria would require the DWS to be designed for high pressure, which is not practicable.

The PSS is connecte-I to the RCS through the local sample points in the RCS hot legs, pressurizer vapor and liquid spaces, and the core makeup tanks. Each of these sampling connection lines contains a flow-restricting orifice that limits the flow from the RCS in the event of a sample line break, and also reduces the pressure in the sampling lines during sampling operations. Each sampling line also contains a normally closed isolation valve before connecting to a common header. The common header then penetrates outside the containment with two normally closed CIVs, which are also the PlVs and will be isolated on a safeguards signalif they are open for sampling operation. The sampling line then connects to a sample cooler and the sample bot *les. In addition, one of the two lines connected to the low-pressure portion of PSS contains two check valves, and the other contains one check valve and one normally closed isolation valve. In the event that these valves leak, the leakage would not overpressurize the low pressure portions of the system, but would tiow directly to the EWST. In the unlikely event of a gross failure of the high pressure check valves, the maximum flow rate from the RCS would be within the capability of the normal makeup system. The water level in the EWST is monitored, and a high alarm in this tank would alert the operator to a potential leak into the tank fiom the PSS uampling lines. The operator would then be able to isolate the leak by closing the CIVs. The CIVs have remote position indication in the control room and are subject to the CIV leakage test.

Therefore, the PSS design meets the intent of the ISLOCA criteria with the exception of PlV leak test requirements the CIVs, which are also PlVs. These PlVs should be included in TS 3.4.16 for PlV LCO and surveillance requirements This is an Open item.

4 4

E 14 ,

(e) Solid Radwesta System:

The nild radweste system (W88), which provides storage facilities for both wet and dry solid wastes prior to and subsequent to processing and packaging, is connected to the f  ; high pressure CVS dominertlisere to facilitate transfer of the spent resin from the CVS

- domineralisers to the spent resin storage tanks (SRSTs). The spent resin header connects to each of the three CVS domineralizers with an individual, normally clo6ed isolation valve, Ed then penetrates containment with two normally closed locked-closed CIVs to the SRSh outside. A manual valve is placed downstream of the outhoard CIV to isolate the downstream piping to facilitate Cly leak testing. The portion of piping downstream of the manual isolation valve is a low-pressure design with the URS below the RCS operating pressure. Section 3.5.2 of WCAP 14428 contends that it is not prectical and necessary to design the WSS to a higher lesign pressure because 6m system contains many low-pressure components such as the SRST and resin transfer and mixing pumps.

The WSS spent sosin line is normally isolated by locked cloor.d manual '.iVs, which are -

administra.tively controlled, have position indications in the control room, and are leak-tested in accordance with the inservice testing plan of SSAR subsection 3.g.8. The CVS domineralisers are inside containment and normally circulate reactor coolant at RCS pressure. Rosin transfer operations are conducted only during refueling operations when

- - - . the RCS is fully depressurised.- During normal power operation, the only pathway to the low pressure portion of trw WSS is for all three closed isolation valves to fal'. Should that .

extremely unlikely event happen, the recirculation loop isolation valves can be closed to isolate the purification loop and the WSS from the RCS In addition, downstream of the inboard CIV in the resin transfer line, there la a relief valve which discharges in the WLS c.'ntainment sump inside containment. Therefore, the WSS spent resin lines are not

required to be designed to a higher design pressure.

. (f) Domineralized Water T ansfer and Storage Systemt The domineralized water transfer and storage system (DWS) receives water from the det..ineralized water treatment system, and provides a reservoir of dominerallaed water to

-supply the condensate storage tank and for distribution throughout the plant. The design and functional details of the DWS is provided in Subsection g.2.4 of the SSAR. The dominersilaed water transfer pumps take suction from the domineralized water storage

. tank (DWST) and supply water through a catalytic oxygen reduction unit to the domineral-laed wuter distribution header. Fr .m this header, domineralized water is supplied to

~

various systems in the plad C..e DWS supply line penetrates containment to a supply r header inside containment, which provides interface of the DWS interface with the PSS and the CVS domineralizers. The DWS provides dominerallaed water to the PSS to flush the

. PSS lines prior to RCS sampling, and to the CVS domineralizers to sluice resin to the WSS.

- The DWS is a low-pressure system design with the ljRS below the RCS operating pressure.

However, the only possible overpressurization pathways from the RCS are the connections

? to the PBS and the CVS domineralizers inalde containment. Overpressurization of the DWS

~

can only eccur if there are multiple failures and misalignments of isolation valves and check valves in the high-pressure systems. A relief valve has been added to the DWS e

na - - - - - _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ - _ - _ . . _ _ . _ . - _ . _ _ - . _ _ _ _ - - - _ _ _ - _ . - -

1 4 0 15 l header inside containment to preclude the possil:;lity of overpressurlaing the DWS. In addition, an overpressurisation of the DWS would most likely result in the rupture of the DWS header inside containment, and therefore is not a concern of 18LOCA.

The staM concludes that, with the exception discussed below, the AP600 design is l consistent with the staff position discussed in SECY-90 016 regarding ISLOCA. Therefore, issue 105 is resolved subject to the resolution of the following open items:

- 0120.3105.1, All PlVs in the RNS, CV9, and PSS should be included in AP600 TS 3.4.16 '

l and suoject to periodic PlV leak testing. This open item is documented in an NRC letter to  ;

Westinghouse (W. Huffman to N. Liparulo), "Open Mem Associated with AP600 3ressure .  !

Isolation Valve Leak Testing," October 16,1997.

0120.3105.2, the relieving capacity and setpressure of the CVS makeup system relief valves, as well as their bases should be documented, lasue 122.2: Initiating Feed and Bleed As discussed in NUREG 0933, Issue 122.2 investigated the findings of the NRC inspection in 1985 of the loss of feedwater event at Davis Besse on June 9,1985. The issue dealt with_.

the adequacy of emergency procedures, operator training, and available plant monitoring systems for determining the need to initiate feed and-blaed cooling following the loss of the SDG heat sSk (i.e., loss of fee.Jwater). In an analysis of the loss of.foedwater event, the staff found that operators weis hesitant to initiate feed and-bleed operations, and that the control room instrumentation was inadequate to alert operators to the need to initiate i feed and bleed. A loss-of feedwater in combination with a failure to diagnose and take corrective actions (i.e., initiate feed and bleed) would result in loss of core cooling.

The staff has completed its review of Westinghouse provided Information relating the feed and blood emergency guidelines AFR H.1, "AP600 Response M Loss of Heat Sink," and has concluded that the feed and blood emergency guidelines are acceptable. - Therefore, this issue is resolved for ths AP300 design. Open item 20.318 is closea, lasue I.D.5(3): Control Room Design - On-Line Reactor Surveillance Systems As discussed in NUREG 0933, issue 1.D.5(3) addressed the benefit to plant safety and operations of continuous on-line automated surveillance systems.- Systems that automati-cally monitor reactor performance can benefit plant operations and safety by providing

- continuous diagno.rtic information to the control room operators, to predict anomalous plant behavior.

Various methods of on-line reactor surveillana have been used, including neutron noise-monitoring in BWRs to detect vibrations in interrial compoaents, and pressure noise surveillance at TMI.2 to momtor primary loop degasification. On-line surveillance data have been u..ed to assess loose thermal shields.

16 .

Continuous on-line surveillance of the NSSS Involves the following areas for which accep-taneo criteria are separately defined:

• vibration monitor!ng of reactor Internals

• RCPB leakage detection e loose parts monitoring The acceptance criteria for the resolution of 1ssue i.D.5(3) for monitoring vibrations in Internal components are in ANSI /ASME OM-51981," Inservice Monitoring of Core Support Barrel Axial Preload in Pressurized Water Reactors." This standard makes recommerida.

tions on the use of ex core neutron detector signals for monitoring core barrel axial preload loss. This standard also documents a program containing baseline, surveillance, and diagnostic phases and makes recommendations for data acquisition frt:lue. icy and analysis.

The acceptance triteria for leak monitnring are in RG 1.45 that documents acceptable methods for channel separation, leakage detection, detection sensitivity and response time, signal calibration, and seismic qualification of RCPB leakage detection systems, it defines the regulatory position for an acceptable design of these systems.

The acceptance cr;teria for loose parts monitoring are in RG 1.133," Loose Part Detection Program for the Primary System of Light-Water-Cooled Reactors." This RG gives gulde-lines an such system characteristics as sensitivity, channel separation, data acquisition, and selsmic and environmental conditions for operability. It also identifies alert levels, data acquisition modes, safety analysis reports, and TS pertaining to a LPMS.

AP600 design includes the reactor coolant pressure boundary leakage detection system as required by 10 CFR 50, Appendix A, General Design Criterion 30, and the conformance to the staff regulatory posNons, as indicated in RG 1.133, for the design of the loose-parts monitoring system. The detalled system design discussions are in SSAR Chapters 517.

The staff has reviewed the functional requirements of the metalimpact monitoring system (MIMS), which monitors the reactor coolant system for the presence of loose metallic parts according to the regulatory position requirements as indicated in RG 1.133, rav 1, May 1981, and concludes that the MIMD functional design requirements satisfied RG 1.133, with the exceptions of system surveillance and reporting requirements, which are most appropriately addressed by the Combined Operating Licensees for plant specif;c design.

Issue 1.D.5(3) is resolved.

Issue ll.D.3: Coolant System Valves Valve Position Indication As discussed in NUREG-0933, Issue il.D 3 addresses the requirements in NUREG 0737 for positive indication in the control room of RCS relief or safety valve position. The accep.

, tance criterion for the resolution of this issue is that the plant design shallinclude safety and relief valve indication derived from a reliable valvo-position detection device or a l

I

m 4

17 l reliable indication of flow in the discharge pipe in accordance with the requirements in NUREG 0737. This indication shall have the following design features.

e Unambiguous safety and relief valve Indication shall be provided to the control room operator.

l e Valve position should be indicated within the coittrol room and 'should be alarmed.

e Valve position indication may be either safety or control gir : if it is control grade, it must be powered from a reliable (e.g., battery backed) Instrument hus (see RG 1.97).

l e ..

Valve position indication should be seismically qualified consistent with the compo-i nont or system to which it is attached.

e Valve position indicatior, shall be qualified for the appropriate operating environment which includes the expected nor ,.al containment environment and an OBE.

o Valve position indication sha'l be human-factors engineered, As discussed in the staff AP600 DSER, confirmatory item 20.41 requires that Westing-house update SSAR Table 3.11 1 to include remote positive indication for the pressuriser safety valve, normal RHR relief valve, and steam generator safety valves. The staff has reviewed the latest SSAR revision 9, Table 3.11 1, which indicates that positive indications have been included for these valves, item 20.41 is considered closed. TMI action item II.D.3 is resolved.-

Issue ll.E.2.2: Research on Smal Break LOCAs and Anomalous Transients As cilaciissed in NUREG 0933, lasue ll.E.2.2, addressed the NRC research programs focused on small Nreak LOCAs (SBLOCAs) and reactor transients. The programs included experimental research in the loss of flow tests (LOFT), semiscale LOFT, Babcock and Wilcox integral systems test facilit8:s, systems engineering, and material effects programs, as well as analytical methods development and assessments in the code development program. - .

The piograms called for in this issue were completed by the NRC and showed that ECCSs will provide adequate core cooling for SBLOCAs and anomalous transients consistent with the ningle-fallure criteria of Appendix K to 10 CFR Part 60. The application of the experi-mental data from the research programs to validate the conservatism of the licensing coded used in the SBLOCAs are addressed in issue ll.K.3(30) in this section. -

Westinghouse did not address this issue in its May 28,1993, letter. It concluded, in Table 1.9-2 of that letter, that this issue was not relevant to the AP600 design because tills l lasue was resolved with no new requirements.

Becau'se the AP600 design is the first passive advanced LWR design to be reviewed by NRC, the staff is considering how the research for the non passive LWRs apply to this-

. u

i l .

i .

Lj 18 ,

h design. Tho' distinguishing feature of the AP600 lo a dependence on safety systems whose operation is driven by natural forces , such as gravity and stored mechanical energy. t j.i

While passive systems may be conceptually simpler than conventional active systems, they may be potentially more susceptible to system Interactions that can upset the balance j of forces upon which the passive systems depend on for their operation, it should be noted that these " passive" systems still rely one some active operation to place them in  ;

i operation.

I For a design with passive safety systems and without a prototype plant that will be tested

over an appropriate range of normal, transient, and accident conditions, the following ,

requirements, the following is required by 10 CFR 52.47(b)(2)(1)(a):

t

e. The performance of each safety feature of the design has been demons" rated

through either analysis, appropriate test programs, experience, or a cor ibination ,

i thersof.

[ e' interdepender.t effects among the safety features of the design nave been found i acceptable by analysis, appropriate test programs, experience, or a combination 4 thereof. ,

i

!'i e Sufficient data exist on the safety features to the design to assess the analytical tools ,

used for safety and analyses over a sufficient range of normal operating conditions, .

transient conditions, and specified accident sequences, including equilibrium core condRions. i i

Westinghouse has developed test programs for the AP600 design to investigate the l

+

passive reactor and containment safety systems, including component pnenomenological 4 (separate effects) test, and Integral systems tests. The staff has completed and docu-

,. . mented its review of the AP600 testing programs in Chapter 21 of the AP600 FSER. lasue it.E.2.2 is considered closed, issue ll.E.5.1i Design Evaluation As discussed in NUREG-0933, issue ll.E.5.1, addressed the reqwoment for B&W licensees to oropose recommendt.tions on hardware and procedural changes relative to the need for .

methods for damping primary system sensitivity.to perturbations in the once through SG.

In 10 CFTt 50.34(f)(20(xvi), it is stated that a design criterion should be established for the allowable number of actuation cycles of the ECCS and RPS consistent with the expected occurrence rate of severe overcooling events considering anticipated trans! ants and accidents.

' Westinghouse identified in Section 1.g.3 of thie SSAR that it considered issue ll.E.5.1 relevant to the AP600 design and stated that although this issue applies only to B&W designs, the AP600 design uses the passive core cooling system to provide emergency reactor coolant inventory control and emergency decay heat removal. Component design criteria has been established for the number of actuation cycles for the passive corw

~ ~ . _ __._____ - - - -

f  !

, i i ..

l 19 i cooling system. The identified actuation cycles include inadvertent actuation, as well as  !

!- the system response to expected plant trip occurrences, including overcooling events.-  ;

[ Operatiots of the ADS is not expected for either design basis or best estimate overcooling events. Section 3.9.1 of the SSAR has additionalinformation.

)- 'The staff reviewed Table 1.9 2, which provided status of TMl and USl/GSl related items

discussions, including item II.E.f.1 In Section 1.9 of the SSAR. The staff considers Open l ltem 20.4-15 closed. >

L ,

} lasuellF.2: Identification of and Recovery From Conditions Leading to inadequate Core Cooling ,

As discussed in DSER Open item 20.417,10CFR 50.34(f), Additional TMI-related Require. ,

monts, requires that instruments be provided in the control room, which have unambigu.  ;

ous indication of inadequate core cooling (ICC), such as primary coolant saturation meters [

L in PWRs, and a suitable combination of signals from Indicators of coolant levelin the  !

L reactor vessel and in-core thermocouples in PWRs and BWRs. NUREG 0737. TMl action i 1

plan item II.F.2, discusses the ICC phenomena and the need to have a reactor water level i I

indication system that provides indication of reactor coolant void fraction when the reactor

{- coolant pumps (RCPs) are operating, and reactor _ vessel water level when the RCPs are ,

j tripped.

[ Prior to the TMI accident, an accepted operational practice of PWRs was to operate the 3 RCPs, if they were available, during a LOCA to provide continued core cooling. During the TMl LOCA event with the stuck open PORVs, the reactor coolant continued to leak through  !

' the opeu valves, the pressuriser level indicated high, and subsequent lCC occurred i because the reactor coolant was highly volded. Nevertheless, core cooling was mahtained i - with the continued operation of the RCPs. Subsequently, the RCPs were tripped and -

i because of high void content in the coolant, the water level dropped below the top of the  ;

core causing fuel damage. As a msuit of the TMI lessons learned, the reactor vessel water

level indication system was added, specifically for PWRs, to ensure operator action to trip the RCPs following a LOCA, rather than later in the LOCA sequence to prevent ICC event, NUREG/CR-5374, Summary of inadequate Core Cooling instrumentation for U.S. Nuclear

Power Plants discusses acceptable approaches to instrumentations used to address ICC.

• 1 in response to stof RAI #440.162, Westinghouse explained that the AP600 design concept ,

- is different from current opemting plants in that the AP600 design automatically trips the -  !

RCPs and initiates safeguard injections through the passive safety systems such es CMT,- .l

- ADS, PRHR and IRWST to maintain core cooling in the event of a SBLOCA. It does not rely on a reactor vessel level indication system as do existing reactors, where reactor vessel -

level indication is important for operator actions to trip the RCPs, to monitor coolant mass p in the vessel and to manually depressurire the RCS in the event of ICC. There is no need In the AP600 for the operetor to trip the RCPs, to inject water into the core or to manually ,

depressurire the plant during a SBLOCA.

[.

4

-The instruments typically used in current PWRs include subcooling margin monitoring

capability, core exit thermocouples, and reactor vessel level indication system, which ,

l1 c

l' I I ._ . - .- L._..._-- - - - . . _ _ _ . . . . . _ _ . _ - . . . . ~ . ... . .,- _.. . ..__ - . _m, .- . _ _ _ . , __ u--.-,,~.

e 20 ,

together would previde the operator with the ability to monitor the coolant conditions and to apf repriately take actions to ensure core cooling during the approach to and to recover from the inadequate core cooling conditions. The AP600 design includes subcooling i margin monitoring capability, core exit thermocouples and the hot leg level indication system. The AP400 hot leg level Indication system is different from the reactor vessei 9, vel indication systems currently use:f in Westinghouse plants.

The AP600 hot-leg level indication is a safety related level indication system, which

! consists of separate pressure taps thr.t connect to the bottom of the hot leg, and to the top of the hot leg bend leading to the steam generator and has the ability to provide Indication of reactor water vessel level for a range spanning from the bottom of the hot leg to approximately the elevation of the vessel mating surface, in addition, during the operation of the ADS to depressuriae the plant, the reactor vessel water level will vary greatly and will not provide a reliable indication of ICC. The AP400 hot leg water level indication is not used to direct operator actions even when the water level may potentially drop below the hot leg level. Therefore, the water level is. not an important Indicatiun for mitigation of ICC in the AP600 design. The hot-leg level indication system is used, however, as i verification of reactor water inven'ory to terminate the recovery action in the ERGS for the ICC event.

Because the AP600 design automatically trips the RCPs during a SBLOCA event and

- because the operators are not prone to be mistaad by forced two phase flow, the core exit temperature is an important and sufficient indication of an approach to ICC condition. The temperature reading provided by core exit thermocouples has been appropriately included in the ERGS for plant recovery.

The staff has reviewed the Westinghouse response and has determined that for a SBLOCA event a safeguard signal would automatically trip the RCPs, passive safety systems st.ch as the CMT would automatically inject water into the core, the ADS would ausomatically initiate to depressuriae the plant, the reactor coolant would automatically be cooled by the PRHR, and subsequent injection from the IRWST would occur. The staff has also dertemined that for AP600 design, the core exit thermocouples and the subcooling martin-monitoring together would provide unambiguous indication of an approach to ICC and the safety related hot leg level indication is only used to terminate the recovery action in the ERGS for the ICC event. Therefore, the requirements fer ICC, as discuss in 10CFR 50.34(f),

have been satisfied and the issue is resolved.

Issue'll.K 1(3): Review operating Procedures for Recognizing, Preventing, and Mitigating Void Formation in Transients and Accidents As discussed in NUREG 0933, issue ll.K.1(3) requested licensees to have operating proceduies for recognizing, preventing, and mitigating void formation in the RCS auring transients and accidents to avoid loss of the core-cooling capability during natural

..c!rculation.

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The staff has reviewed the resolution of issue 1.C l and its related ERGS AES 0.2, " Natural

Circulation Cooldown," and has concluded the guidelines direct the operators to cooldown  !

j and depressurize the plant using natural circulation conditions by dumping steam and li subsequent RNS operation. These steps are specified to preclude any possible upf er head voids formatin.f and also direct the operators to verify that a steam vold does not exit i

{ in the vessel. T_he staff concludes that the ERGS provide directions to plant operators to  ;

, recognize and _to preclude volds formation in the vessel and therefore, the staff considers j_ lasue ll.K.1(3) closed, issue ll.K.1(4d): Review operating Procedures and training to Ensure that Operators are Instructed Not to Rely on Level Alone in Evaluating Plant Conditions - j As discussed in NUREG 0933, issue ll.K.1(40) asked licensees to provide operating  !

! procedures to ensure that operators shall not rely on level indication alone in evaluating [

plant conditions.~ As stated in NUREG-0933, the staff determined that this issue was  ;

covered by issues 1.A.3.1, l.C.1, and ll F.2, and is resolved.  :

lasue I.A.3.1, " Revise Scope and Criteria for Licensing Examinations," was implemented by NRC by a rule change to 10 CFR Part 55, " Operators Licenses," to require simulator as part '

of the reactor operator liceneing examinations. The staff w!Il impose the requirements of 10 CFR 55.45 on simulators on the COL applicant referencing the AP600 design; therefore, Westinghouse and the staff does not have to address issue I.A.3.1 for compliance with q 10 CFR 52.47(a)(1)(lv). i Westighoung did not addrets this issue in its May ?A.1993, letter, it concluded, in

-Table 1.V 2 of that letter, that this issue was r.ot relevant to the AP600 design because Ols  ;

- issue is not a design certification issue, but is the responsibility of the COL' applicant.

However, in response to the staff request for additional information (RAI), Westinghouse stated that the design portion of this item is addressed in the proposed resciution to lasues 1.C.1 and ll.F.2.

The staff has completed its review of issues I.C.I and ll.F.2 and has concluded that AP600 (

ERGS do not inste uct the operators to rely on level indication alone in evaluating plant '

conditions. The status of core cooling is determined by indications of core exit thermocou.

pie temperature, RCS subcooling, and RCS hot leg temperature in addition with RCS level.

The staff cons,iders these issues resolved and therefore, issue ll.K.1(4d) is clcsed, issue ll.K.1(17): Trip Pressurizer Level Bistable so that Pressurizer Low Pressure Will I initiate Safety injection As discussed in the staff DSFR Open item 20.4 22, TMl action plan item II.K.1(17) ad.

dresses the requirement for Westinghouse plants to trip the pressurizer level bistable so that the pressurizer low pressure, rather than the pressurizer low pressure and pressurizer low level coincidence, would initiate safety injection.  !

AP600 design does not depend on pressurizer low pressure and pressurizer low level coincidence to initiate safety injection in the event of LOCAs. Safety injection in AP600 l

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design is automatic. The following safeguard signals would initiate safety injee' ion: Low.

B 1 pressuriser pressure or Hi 1 containment pressure or Low compensated steam line i pressure or Low 3 cold leg temperature. In addition, the t?600 design also gives the

. operator manual safety injection capability. The staff concludes that any single safeguard

[ signals mentioned above would initiate safety injection. Therefore, this issued is resolved.

4 .

t i issue li.K.1(24): Perform LOCA Analyses for a Range of small-Break Slees and a Range of j i - Time Lapses EJoiween Reactor Trip and RCP Trip i

l losue ll.K.1(24), of NUREG4933 required PWR licensees to perform a LOCA analysis for a l range of small-break slaes and a range of time lapses between reactor trip and RCP trip.

i - The staff determined in NUREG-0933 that this issue for PWRs was covered by issue I.C.1, 2

"Short Term Analysis and Procedures Revision."

Westinghouse has provided for staff review of AP600 Emergency Response Guidelines (ERGS), which addresses lasue 1.C.1. The staff has reviewed the responses to issue I.C.1 and specifically the emergency response guideline AE 0, "AP600 Reactor Trip or Safety t

! Injection" for small break LOCA that addresses item II.K.1(24) and has concluded that the r AP600 design automatically trips the RCPs during a LOCA event. The guideline directs  ;

i the operators to verify tnat all reactor coolant pumps have been tripped, and if not, the  !

! operators are directed to manually trip the reactor coo' ant pumps. Based on the plant l l . design features and the appropriate operator's actions using ERGS, the staff considers ,

item II.K.1(24) resolved. Open item 20.4 23 is closed. ,

I

!- Issue ll.K.1(25): Develop Operator Action Guidelines f l As discussed in NUREG 0933, issue li.K.1(25) required PWR licensees to develop operator ,

action guidelines based on the analyses performed in response to lasue ll.K.1(24), which is  !

- discussed above. -The staff determined in NUREG 0933 that this issue was covered by issue I.C.1.

o Westinghouse did not address this issue in its May 28,1993, letter. It concluded, in L Table 1.9 2 of that letter, that this issue was not relevant to the AP600 design because the . 1 issue had been superseded by one or more other issues. Although this issue was covered .

- by issue I.C,1, as stated above, Westinghouse also did not address this latter issue j because it considered issue I.C.1 the nie responsibility of the COL applicant.

The final procedures would be the responsibility of the COL applicant; however, the range of LOCA analyses for a range of time lapses and the specific information to go into the .

procedures would be the responsibility of the designer, or Westinghouse in the case of the 1

[ AP600 design. Westinghouse addresses accidents for the AP600 design in SSAR

. Chapter 15. The staff requests that Westinghouse address operator action guidelines, or i EPGs, of I.C.1 and the role of the COL applicant in lasue ll.K.1(25). This is Open  ;

Item 20.4 24.

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The staff has completed its review of issue I.C.1 and has conclude:I that lasue I.C.I is closed, therefore, Issue ll.K.1(25) or Open item 20.4-24 is also closed.

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Issue ll.K.1(27): Providc Analyses and Develop Guidelines and Procedures for inadequate Core Cooling a

As discussed in the staff DSER, the AP600 design should describe analyses of ICC conditions and develop guidelines and procedures to mitigate an ICC event, and that this

. Issue is dependent on the resolution of Action items ll.F.2 and I.C.1, Identification of and Recovery from conditions Leading to ICC and Guidance for Evaluation and Development of Procedures for Transients and Accidents, respectively.

Westinghouse has submitted AP600 Emergency Response Guideline (ERG) for staff  !

rsview, and also responded to staN DSER Open item 20.417 to address Action items I.C.1 1

- and ll.F.2, respectively. The staff has reviewed Westinghouse response to Action item 3 II.F.2 and a detailed discussion of this item is documented in its respective section. In the i

). AP600 ERG, Westinghouse provides high-level guidance to deal with inadequate core l cooling conditions The staff has reviewed AFR C.1, AP600 Response to inadequate Core i j Cooling procedure and analysis bases, which describss how passive safety-related systems would automatically trip the RCS pumps, initiate and depressurize the RCS to l inject water into the core upon receiving a safeguard signal, in this procedure, the

operators are instructed to monitor plant conditions using core exit temperature and I

Indicated hot leg level, which is designed to provide Indication of an approach to ICC and  !

! to recover from an ICC condition. The operators are also instructed to manually initiate

injection when automatic passive safety injections fall. Passive safety related system ,

, actuation Indications of CMT, ADS, PRHR, and IRWST are integrated into the procedures,  ;

i which provide operators with directions to ensure that adequate core cooling will be . .

4

' maintained. Therefore, the staff concludes that Westinghouse has appropriately provided l

! analyses and procedures to mitigate ICC conditions, issue ll.K.1(27) or Open item 20.4 25 1 is closed. <

4

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j issue ll.K.3(6): Instrumentation to Verify Natural Circulation ,

As discussed in NUREG 0933, Issue ll.K.3(6), addressed requiring licensees to provide-E ~

mstrumentation to verify natural circulation during transient conditions. The staff deter. -

mined in NUREG 0933 that this issue was covered by issues I.C.1, ll.F.2, and ll.F.3.

1-I Westinghouse has provioed the staff with pertinent information about the AP600 design, 1 which addresses TMl action items I.C.1, ll.F.2 and ll.F.3. . The staN has reached a conclu-slon that those issuc. relevant to the resolution of the TMI action item II.K.3(6) have been i resolved. The detailed discussion of the related issues are addressed in their respective TMlitem discussions. Therefore, this issue is resolved.

i lasue ll.K.3(8): Further Staff Consideration of Need for Diverse Decay Heat 11emoval g Method Independent of Steam Generator As discussed in NUREG-0933, issue ll.K.3(18) addressed further sta# consideration of the need for diverse decay heat removal methods which were independent of the steam

- generators. The staff determined in NUREG-0933 that this issue was covered by issues _,

ll.C.1, " Interim Reliability Evaluation Program," and ll.E.3.3, " Coordinated Study of Shut-  ;

I

24 l down Heat Removal Requirements." In NUREG 0933, the stan also stated that issue ll.E.3.3 l

was addressed in lasue A 45," Shutdown Decay Heat removal Requirements."

l Westinghouse has prov8ded the AP600 shutdown evaluation report for staM review. The L report describes multiple decay heat removal capabilities independent of the steam _

! generator. The detailed discussion of the multiple decay heat capabilities is included in Chapter 19.3 of the FSAR. -The sta#, therefore, concludes that Issue ll.K.3(8) or Open item 20.4 27 is closed.

l losue ll.K.3(30): Revised SSLOCA Methods to Show Compliance with 10 CFR Part, Appsn.

dix K As discussed in NUREG 0933, lesue ll.K.3(30) required licensees to revise and submit analytical methods for small-break LOCA analyses for compliance with Appendix K to 10 CFR Part 50 for NRC review and approval. The revision was to account for comparisons with experimental data, including data from LOFT test and semiscale test facilities.

Altematively, licensees were to provide additional justification for the acceptability of their SBLOCA models with LOFT and semiscale test date. Clarifications were issued in NUREG 0737. The staM has reviewed NOTRUMP code and has documentec8 lts discus.

sions in Chapters 15 and 21 of the 78ER. The stan, therefore, considers this issue closed.

.

b 25 BL/GL STATUS BL/GL TITLE DSER RESOLUTION FSER RESOLUTION BL-80-12, Decay heat removal This bulletin dealt with re- The staff evaluates this issue ducing the likelihood of los- in FSER Section 6.3. This ing DHR capability. W GL is CLOSED.

stated that this issue is dis-cussed in SSAR 7.4.1 BL 80-18, maintenance of ade. W stated the design do not N/A. CLOSED.

quale mini flow through CCP have CCP as part of Sl and following secondary side high that it is not applicable to J energy line rupture AP600.

BL 86-01, mini flow logic prob- AP600 does not have valves N/A. CLOSED.

lem that could disable RHR in mini flow lines. Issue re-pumps solved BL 89-03, potentialloss of re- The staff indicated that SSAR Section 9.1 discusses quired shutdown margin during movement and placement of fuel storage and handling refueling fuel during refueling is within including the refueling equip-the scope of AP600 core ment which is used to safely design. This issue is a COL move and store fuels. Addi-action item tionally, IRWST provides large quantities of borated water that rnaintains the re-quired shutdown margin.

during refueling. The BL 89-f 03 issue is CLOSED and the COL action item is still valid regarding plant specific

__ guidelines.

GL-80-01, report on ECCS The staff requested that W The safety evaluation of th s address NUREG-0630, issue is in FSER Chapter 15.

" Cladding, Swelling and The GL-80-01 is CLOSEU.

Rupture Models for LOCA Analysis."

GL 80-014, LWR primary W needed to indicate where W stated that SSAR 1.9.4.1.2 coolant system pressure in Section 1.9 of the SSAR and USi-B-63, discuss this isolation valves that this issue is discussed issue. The staff discusses this issue in the ISLOCA context and is having an outstanding open issue (TMl/ Issue 105). This is an OPEN ITEM.

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GL 80-019, resolution of W stated that the fission gas The staff position ha not enhanced fission GPS felease release models are changed. '.) action is concern accounted for in WCAP- needed. The GL-80-019 is '

10851 P A and WCAP- CLOSED.

11873 A," improved fuel performance models "or W fuel rod design and safety evaluations " This is, sue is resolved GL-81021, Natural circulation The staff stated that W W has submitted AP600 cooldown should address the ERG for ERG GW GJR 100, Rev 3 this event. dated 5/97 for staff review.

The staff has reviewed this submitta' and its related natural circulstion ERG and has determined that guidelines are sufficiently given to the operator to cool down the plant using natural circulation ineans This GL is CLOSED.

GL 83-11, licensee W should address the This issue is COL qualifications for performing quahfications for performing responsibility. This GL is safety analyses in supporting safety analysis for AP600 CLOSED.

licensing actions design.

GL 64 21,long term, low- Core peaking factor may be The safety evaluation of this power operation in PWRs greater than assumed in issue is in FSER Chapter 15.

safety analysis for extended This GL is CLOSED.

Iow power operation following a return to full ptwer ops.

GL 85-16, high boron This GL is resolved because The staff position has not concentratiori AP600 design does not changed. No action is have BIT and born required. Thic GL is concentration from CMT is CLOSED.

much lower than BIT (22,000 ppm).

GL 8712, loss of RHR with This GL addressed potential W has submitted WCAP-RCS partially filled, forloss of RHR during 14837, Rev 2 (11/97) that midloop operation discusses shutdown risk concerns, including potential loss of RNS. The staff has resolved this issue and its SE is discussed in FSER chapter

19. This GL is CLOSED.

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e 27 GL 8817, loss of DHR This GL addressed W has submitted WCAP-potentialforloss of RHR 14837, Rev 2 (11/97) that during midloop operation discusses shutdown risk concems, including potential loss of RNS. The staff has resolved this issue and its SE is discussed in FSER  !

Chapter 19. This GL is l CLOSED.

GL 9107, P.' ** seat failures GSI 23 discussed RCP The staff has resolved GSI-  !

pump seal failures. W 23 issue because W design '

addressed this issue in does not have RCP pump SSAR Sections 5.1.3.3, and seals, and the GL is not 1.9.4.2.3 applicable to AP600 canned pump design. This GL is CLOSED.

GL 9304, rod control system W should revise the WCAP- WCAP has been revised failure 13559 to include this ilum of (8/96) to include reference of discussion for AP600 design this item of discussion in SSAR 3.9.4. The staff SE is discussed in FSER Chapter

4. This GL is CLOSED.

GL-83-22, safety evaluation of THIS ISSUE WAS NOT The staff has reviewed W ERGS INCLUDED IN DSER AP000 ERG-GW GJR 100, Rev 3 dated 5/97 and has documented its evaluation in FSER Section 18.9.3. This GL is CLOSED.

GL 86-07, NUREG 1190 _

THIS ISSUE WAS NOT This issue is not part of regarding the San Onofre Unit INCLUDED IN DSER SRX3 responsibility and it is 1 loss of power and water most appropriately addressed hammer by HHFB.

28 *.

BL 96-01, rod control problem THIS ISSUE WAS NOT The BL was issued because INCLUDED IN DSER of incomplete control rod insertion (IRI) evaluation at the South Texas and Wolf Creek plants. it has been determined that the IRI was

, caused by the thimble tube distortion resulting from

excessive load. Since this is a fuel design problem, and W has not committed to any fuel types and tnat this problem is mostly resolved by the fu-manufacture"s, the staff '

,

I concluded W does not have l to address this issue, unless it has committed to certain fuel designs discussed in the BL. This issue should be appropriately addressed by the COL applicant. This is a new COL action item. The BL is CLOSED.

GL 8616, ECCS evaluation The staff requested that W W discussed this issue in models discuss the ECCS SSAR Sections 6.3.5 and evaluation models for 15.0.11. The staff has AP600 design. evaluated Westinghouse ECCS models and has discussed this istoo in FSER Chapter 15. This GL is CLOSED.

GL 85-05, Inadvertent Boron W discussed this issue in Dilation SSAR Section 15.4.6. The staff has evaluated and has discussed this issue in FSER Chapter 15. This GL is CLOSED.

GL-96-04, Boraflex and Spent This issue is not part of Fuel Racks SRXB rosponsibility. It is most approprietely address,ed by SPLB.

Note: All TMl items ( 11.B.1, ll.G.1, ll.K1(28), ll.K.2(16), ll.K.3(2), ll.K.3(5) and ll.K.3(25) indicated in the 12/10/97 note from J.Sebrosky have bee:e resolved in the DSER. _ The staff does not see any changes in position regarding its evaluation of these items. Also, issue 23 has been resolved and is reflected in DSER. No change in the staff position is anticipated.

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