NSD-NRC-98-5507, Forwards W Responses to FSER Open Items on A600.Summary of Encl Responses & Markup of PRA App D Also Encl

Forwards W Responses to FSER Open Items on A600.Summary of Encl Responses & Markup of PRA App D Also Encl

ML20198J409
Person / Time
Site:	05200003
Issue date:	01/05/1998
From:	Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NSD-NRC-98-5507, NUDOCS 9801140124
Download: ML20198J409 (110)

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NSD-NRC 98 5507 Docket No.: 52 003 7 January 5,1998 1

Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTEN_ TION: T. R. QUAY-

SUilJECTi AP600 RESPONSE TO FSER OPEN ITEMS AND MARKUP OF PRA APPENDIX D

Dear Mr._ Quay:

~ Enclosure 1 of this letter provides the Westinghouse responses to FSER open items on the AP600. A summary of the enclosed responses is provided in Table 1. Included in the table is the FSER open item number, the associated OITS number, and the status to be designated in the Westinghouse status column of OITS.

Enclosure 2 provides a marbp of PRA Appendix D. This appendix is being revised based on the responses to FSER open items 720.414F and 720.416F. Enclosure 2 shows where text will be

removed when this section is placed in PRA Revision 11. The new informrtion that is provided via the enclosure is shown in italic, boldfaced type.

The NRC should review the enclosures and inform Westinghouse of the stat is to be designated in the "NRC Status" column of OITS.

Please contact me on (412) 374-4334 if you have any questions concerning his transmittal.

} $"0if/

lirian A. McIntyre, Manager f

'I Advanced Plant Safety and Licensing

-jml-Enclosure H004:

cc: W, C. IlulTman, NRC (Enclosure 1)

- T. J. Kenyon, NRC (Enclosure 1)- _

J. M. Sebrosky, NRC (Enclosures 1&2)

D. C. Scaletti, NRC (Enclosure 1)

N. J.: 1.iparuto, Westinghouse (w/o Enclosures) 9901140124 9M105 ll1515115NE ADOCK %20 ,0003 IEI.M.HE.BIMaRum IH. ill R.H.EE.

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DCP/NRCl202 NSD-NRC 98 5507 2- January 5,1998 Table !

List of FSER Open Items included in Ixtter DCP/NRCl202 FSER Open item OITS Number Westinghouse status in OITS 410.398F 6425 Action N 410.399F 6426 Action N 4

650.35 F 6432 Confirm W m

t Enclosure 1 to. Westinghouse Letter DCP/NRC1202 January 5,1998 4

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o FSER Open >n 410.398F (OITS #6425)

Provide pipe break analyses. De response to RAI question 410.76 was an interim response. (O!

3.6.14)

Responw:

A summary of the results of the pipe break analysis may be found in SSAR tables 3.61, 3.6 2, and 3.6 3. Dese results include the information requested in RAI question 416.?d. De calculations that support these pipe break results are included in our engineering files and are available for review by the NRC staff.

SSAR Revision: NONE a

h MDM 410.398 1 i

o FSER Open hem ,

410.399F (OITS #6426)

Provide analysis for the failure of non seismic Category I systems. (OI 3.6.15)

Response

Both seismic and nonseismic high energy piping in the nuclear island is evaluated for adverse effects of pipe breaks on safety related systems and components. 'Ihe criteria for the location of high-energy breaks is provided in subsection 3.6.2.1.1 for seismically analyzed and nonseismically analyzed piping.

'Ihe results of the pipe break evaluation are tabulated in Table 3.6 2 and 3.64. -

Note that piping directly connected to safety related piping or components 1.1 analyzed as seismic Category 119 to at least the first anchor. As a result, there is minimal non:cis.nically analyzed piping that must be analyzed for effects of pipe rupture.

SSAR Revision: NONE 9

UD 410.399 1 i

b FSER Open hem sW.

650.35F: Issue A 9: Anticipated Transient Without Scram (OITS #6432) 10 Crit 50.62 requires the autornatic initiation of the au. Sary feedivater system under conditions indicative of an ATWS. The AP600 does not include this design feature, but instead automatically initiates the PRilR system. The applicant should request an exemption from the 10 CFR 50.62 for automt. tic initiation of auxiliary feedwater, and update the discussion in SS AR Section 1.9.4.2.2 accordingly.

Response

The AP600 meets the intent of the regulation with the use of the passive residual heat removal system.

The SSAR will be revised and an exemption will be requested.

SSAR Revision:

Revise the response to A 9 in subsection 1.9.4.2.2 as follows:

AP600 Response:

I The AP600 complies with the requirements of 10 CFR 50.62 except that the AP600 does not have i a safety related auxiliary feedwater system. In lieu of the automatic initiaticn of the auxiliary I feedwater system under conditions indicative of an ATWS as required by 10 CFR 50.62 (c)(1), the i AP600 automatically initiates the passive residual heat removal system as discussed in Section 6.3.

1 I A discussion of the AP600 design features used to address the probability of an ATWS is presented in subsection 1.9.5 and Section 7.7.

MDM 650.35 1

I:nclosure 2 to Westinghouse Letter DCP/NRCl202 January 5,1998 9

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p r; '; at Servinbeity Assessanut APPENDIX D EQUIPMENT SURVIVABILIYY A3SESSMENT D.1 Introduction ne pu pose of the equipment survivability assessment is to n-luate the availability of equipment and instrumentation used during a severe accident to achieve a controlled, stable state after core damage under the unique containment environments. Severe accident phenomena may create brush, high temperature and pressure containment environments with a significant concentration of combustible gases. Local or global buming of the gases may occur, prese.iting additional challenges to the equipment. Analyses demonstrate that there is reasonable. assurance that equipment used to mitigate and monitor severe accident progression is available at the time it is called upon to perform.

De methodology used to demonstrate equipment survivability is:

• Identify the high level actions used to achieve a controlled, stable state ,

• Define the accident time frames for each high level action a Determine the equipment and instruments used to diagnose, perform and verify high level actions in each time frame

• Determine the bounding environment within each time frame

• Demonstrate reasonable assurance that the equipment will survive to perfomi its function within the severe environment.

D.2 Applicable Regulationss and Criteria Equipment tha. is classified as safety related must perform its function within the environmental conditions associated with design bases accidents, ne level of assurance provided by equipment requ ;cd for design bases events is " equipment qualifica' ion."

De environmental conditiora resulting from beyond design basis events may be more limiting than conditions from design bases events. The NRC has established criteria to provide a reasonable level of assurance that necessary equipment will function in the severe accident enviromnent within the time span it is required, nis criterion is referred to as equipment survivability."

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9' D. Famipawat Sureliability Assessment he applicable criteria for equipment, both mechanical and electrical, r quired for recovery from in vessel severe accidents are provided in 10 CFR 50.34(0:

• Pan 50.34(D(2XixXc) states that equipment necessary for achieving and maintaining safe shutdown of the plant and maintaining containment integrity will perform its safety function during and after being exposed to the environmental conditions attendant with the release of hydrogen generated by the equivalent of a 100 pettent fuel clad metal-water reaction including the environmental conditions created by activation of the hydrogen control system.

• Part 50.34(O(2Xxvii) requires instrumentation to measure con'ainment pressure, containment water level, containment hydrogen concentration, conWnment radiation intensity, and noble gas effluent at all potential accident release points.

• Part 50.34(O(2)(xix) requires instrumentation adequate for monitoring plant conditions following an accident that includes core damage.

• Part 50.34(0(3Xv) starm M ratems necessary to ensure containment integrity shall be demonstrated to perfes w w.ction under conditions associated with an accident that releases hydrogen gea M %m 100 percent fuel clad metal water reaction.

De applicable criteria for equipment, both niectrical and mechanical, required to mitigate the consequences of ex vessel severe accidents is discussed 1.1 Section III.F. " Equipment Survivability" of SECY 90416. De NRC recommends in SECY 93 087 that equipment provided only for severe accident protection need not be subject to 10 CFR 50.49 equipment qualification requirements, the 10 CFR 50 Appendix B quality assurance requirements, or 10 CFR 50 Appendix A redundancyldiversity requirements. However, mitigation features must be designed to provide reasonable assurance they will operate in the severe accident environment for which they are intended and over the time span for which they are needed.

D.3 Definition of Controlled, Stable State ne goal of accident management is to achieve a controlled, stable state following a beyond design basis accident. Establishment of a controlled, stable state protects the integrity of the containment pressure boundary. De conditions for a controlled, stable state are defined by WCAP 13914, the Framewo;t for AP600 Severe Accident Management Guidance (SAMO)

(Reference D 1).

For a controlled, stable core state:

• A process must be in place for transferring the energy being generated in the core to a long term heat sink

• ne core temperature must be well below the point where chemical or physical changes might occur

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D. Equipawat Survitabilky Assessament For a controlled, stable containment state:

• A process must be in place for transferring the energy that is released to the containment to a long term heat sink

• ne containment boundary must be protected

• De containm:nt and reactor coolant system conditions must be well below the point where chemical or physical processes (severe accident phenomena) might result in a dynamic change in containment conditions or a failure of the containment boundary.

DA Definition of Equipment Survivability Time Frames ne purpose of the equipment survivability time frames is to identify the time span in the severe uccident in which specific equipment is required to perform its function. De phenomena and environment associated with that phase of the severe accident defines the environment which challenges the equipment survivability. De equipment survivability time frame definitions are summarized in Table D.41. I D.4.1 Time Frame 0. Pre Core Uncovery Time Frame 0 is defined as ths period of time in the accident sequence after the accident initiation and prior to core uncovery. De fuel rods are cooled by the water / steam mixture in the reactor vessel. Tne accident has not yet progressed beyond the design basis of the plant and hydrogen generation and the release of fission products from the core is negligible.

Emergency response guidelines (ERGS) are designed to maintain or recover the borated water inventory and heat removal in the reactor coolant system to prevent core uncovery and establish a safe, stable state. Recovery within Time Frame 0 prevents the accident from becoming a severe accident. Equipment survivability in Time Frame 0 is covered under the design basis equipment qualification program.

D.4.2 Time Frame 1. Core Heatup Time Frame 1 is defined as the period of time after core uncovery and prior to the onset of significant core damage as evidenced by the rapid oxidation of the core. His is the transition period from design basis to severe accident environment, ne overall core geometry is intact and the uncovered portion of the core is overheating due to the lack of decay heat removal.

Hydrogen releases are limited to relatively minor cladding oxidation and sonie noble gas and volatile fission products may be released from the fuelclad gap. As the core-exit gas temperature increases, the ERGS transition to a red path indicating inadequate core cooling.

De operators attempt to reduce the core temperature by depressurizing the RCS and re-establish the borated water inventory in the reactor coolant system, if these actions do not result in a decrease in core exit temperature, the control room staff initiate actions to mitigate a severe accident by tuming on the hydrogen igniters for hydrogen control and flooding the T M f $ 0 tis 4 hy.,,,

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a D, Equipment SurtivabLlity Assessment reactor cavity to prevent reactor pressure vessel failure. Recovery in Time Frame i prevents the accident from becoming a core melt. Equipment survivability in Time Frame 1 is evaluated to demonstrate it is within the equipment qualification envelope.

D.4.3 Time Frame 2. In Vessel Severe Accident Phase Time Frame 2 is the period of time in the severe accident after the accident progresses beyond the design basis of the plant and prior to the establishment of a controlled, stable state (end of in vessel core relocation), or prior to reactor vessel failure. De onset of rapid oxidation of the fuel rod cladding and hydrogen generation defines the beginning of Time Frame 2. ne heat of the exothermic reaction accelerates the degradation, melting and relocation of the core.

Fission products are released from the fuel clad gap as the cladding bursts and from the fuel matrix as the UO2pellets melt. Over the period of Time Frame 2, the initial, intact geometry of the core is lost as it melts and relocates downwed inside the core reflector. ne, molten corium pool eventually melts through the renector and relocates to the lower head. Severe accident management strategies exercised during Time Frame T, are designed to recover reactor coolant system inventory and heat removal, to maintain reactor vessel integrity and to maintain containment integrity. Recovery actions in Time Frame 2 may create environmental' challenges by increasing the rate of hydrogen and steam generation.

D.4.4 Time Frame 3. Ea Vessel Severe Accident Phase Time Frame 3 is defined as the period of time after the reactor vessel fails until the establishment of a controlled, stable state. De AP600 reliably provides the capability to Good the reactor vessel and prevent the vessel failure in a severe accident, and, as quantified in the PRA, this severe accident time phase 3 is of such low frequency, it is considered to be remote and speculative. Molten core debris is relocated from the reactor vessel onto the containment cavity floor which creates the potential for rapid steam generation, core concrete interaction and non condensible gas generation. Severe accident management strategies implemented in Time Frame 3 are designed to monitor the accident progression, maintain containment integrity and mitigate fission product releases to the environment.

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A D. Equipment Survivability Asaeesment I

Table DA 1 DEFINITION GF EQUIPMENT SURVIVABILITY TIME FRAMES Time Frame lleginning Time Ending Tlrne Comments O Accident safe, stable state

• Bounded by design basis equipment initiation or qualincation environment core uncovery 1 Core uncovery controlled, stable

• Core uncovery and heatup statt

• Bounded by design basis equipment or qualification environment rapid cladding osidation 2 Rapid cladding controlled, stable e in vessel core melting and relocation osidation state

• Entry into SAMO or vessel failure 3 Vessel failure controll(d, stsble . Ex vessel core relocation state or containment failure D.5 Definition of Active Operation Time Equipment only needs to survive long enough to perform its function to protect the containment fission product boundary. In the case of some items, such as valves or motor-operators, once the equipment performs its function, it changes state and the function is completed. For other items, such as pumps, the equipment must operate continuously to perform its function. The time of active operation is the time during which the equipment must change state or receive power to perform its function.

D.6 Equipment and Instrumentation for Severe Accident Management De AP600 emergency response guidelines (Reference D 2) and severe accident management guidance (S AMO) framework (Reference D 1) define actions that accomplish the goals for achieving a controlled, stable state and terminating fission product releases in a severe accident. De high level actions from the accident management framework are summarized in Table D.61 and provide the basis for the acuons considered for identifying equipment.

De purpose of this section is to review ERO and SAMG actions within each of the time frames of the severe accident to determine the equipment and insuumentation and the active operation time in which they are needed to provide reasonable assurance of achieving a w w.eem. SYL_ ""%"& "

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4 D. Equipment Survivability Aueument controlled, stable state. De AP600-specine accident management framework is used to identify the equipment for performing the high level actions.

The Westinghouse Owners Group (WOG) S AMG (Reference D 3) provides the primary input to the selection of the instrumentation used for monitoring the actions. De instrument used to diagnose the need for the action and monitor the response are listed. Instruments to evaluate potential negative impacts are covered under other high level actions in the framework and therefore are also considered for survivability.

De equipment and insuumentation used in each time frame are summarized in Tables D.6 2 through D.6-4.

D.6.1 Time Frames 0 and I . Accident Initiation, Core Uncovery and Hestup Time Frame 0 represents the accident time prior to core uncovery. Time Frame 1 represents the *ime following core uncovery, prior to the rapid oxidation of the core. Aside from potential ballooning of the cladding, the core has not lost its initial intact geometry and is coolable. .

During Time Franes 0 and I, most of the equipment that is automatically actuated will receive a signal to start. However, given a severe accident sequence, some critical equipment does not actuate. From accident initiation until the time of core uncovery (Time Frame 0) the conditions are bounded by the design basis and covered under equipment qualification.

During Time Frame 1, the environment is still within the design basis of the plant and the control room is operating within the Emergency Response Guidelines (ERGS), but the conditions have the potential to degrade. To achieve a controlled, stable state, accident management, via the ERGS,is geared toward recovering the core cooling before the coolable geometry is lost. Failing that, the plant is configured to keep the core debris in the vessel, and mitigate the containment hydrogen that will be generated in Time Frame 2.

D.6.1.1 Injection into the RCS Failure of RCS injection is likely to be the reason the accident has proceeded to core uncovery. Successful injection into the RCS removes the sensible and decay heat from the core, Prior to the rapid oxidation of the cladding, successful RCS injection essentially recovers the accident before it progresses to substantial core melting and relocation and establishes a controlled, stable state. Failure to inject into the RCS at a sufficient rate allows the accident to proceed into Time Frame 2 and the SAMO.

De equipment and systems used to inject into the RCS are the core makeup tanks, accumulators and IRWST (which are part of the passive core cooling system (PXS)), the chemical and volume control system (CVS) pumps, and the normal residual heat removal (RNS) pumps. For non LOCA and small LOCA sequences, depressurization of the RCS is required for successful injection, ovmv.itw.pt it iisoe? D-6 hhw 3 Westingh0058

.4 D. Equipment Sunivabuky Anemment Injection into the RCS in Time Frame I is covered in a number of ERO procedures. ne FR.C l procedure is entered from the Critical Safety Function Status Tree based on high core-exit temperature, and provides a final attempt to recover the core with water. The plant response is monitored using the system Dowrates, RCS pressure, core exit temperature, or RCS piping temperature.

D.6.1.2 Injection into Containment In ERO FR.C.I the operator is instructed to inject water into the containment to submerge the reactor vessel and cool the extemal surface if injection to the RCS cannot be established.

This action is performed at the end of Time Frame 1, immediately prior to entry into the SAMO. Successful cavity flooding prevents vessel failure in the event of molten core relocation to the vessel lower head. Failure of cavity flooding may allow the accident to proceed to vessel failure and molten core relocation into the containment (Time Frame 3)if timely injection into the reactor vessel cannot be established to cool the core and prevent substantial core relocation to the lower head.

De PXS motor operated and squib recirculation valves are opened manually to drain the "

IRWST water into the containment.

ERO FR.C 1 is entered from the Critical Safety Function Status Tree based on high core exit temperature. De plant response is monitored by containment water level or IRWST water level indication.

D 6.1.3 Injection into the Steam Generators in the event of non LOCA or small LOCA sequences, the RCS pressure is elevated above the secondary pressure. Failure of feedwater to the steam generators may be the initiating event for such sequences and recovery of injection to the steam generators may be required, if the steam generators remain dry and the core is uncovered, the tube integtity or hot leg noule integrity will be threatened by creep rupture failure at the onset on rapid oxidation (entry into Time Frame 2). Injecting to the steam generators provides a heat sink to the RCS by boiling water on the secondary side, and protects the tubes by cooling them. Successful stean generator injection can establish a controlled, stable state if the losses from the RCS cm be recovered and mitigated. Failure to inject to the steam generator requires depressurization of the RCS to prevent creep rupture failure of the tubes and loss of the containment integrity at the onset of rapid oxidation in Time Frame 2.

For accident sequences initiated by steam generator tube rupture, the procedures instruct the control room to isolate injection to the faulted steam generator, and to use injection to the intact steam generator in conjunction with steam generator depressurization to cooldown the reactor coolant system and isolate the break. In Time Frame 1, within the FR.C 1, injection to the intact steam generators may be used to re-establish a primary heat sink to cooldown the RCS and a controlled. stable state if the losses from the RCS can be recovered .gid mitigated.

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Failure to inject to the steam generator may lead to a continued loss of coolant to the fantted steam generator and progression to Time Frame 2.

De main feedwater and startup feedwater pumps are used to inject into a pressurized secondary system. If the secondary system can be depressurized sufficiently, condensate, fire water or service water can also be used to inject into the secondary side.

Injection into the steam generators is covered in the ERO in FR.H.l. He guidelines are entered from the Critical Safety Function Status Tree based on low steam generator level, either wide range or narrow range. ERO FR.C.I is entered based on high core-exit temperature. De plant response is monitored with the steam generator level and steamline pressure.

D.6.1.4 Depressurize Reactor Coolant System D.6.1.4.1 Non LOCA and Small LOCA Sequences in the event of non LOCA or a small LOCA sequences, the RCS pressure is above the secondary pressure. If the steam generators are dry and the core is uncovered, the hot lei nozzle or tube integrity is threatened by creep rupture failure at the onset of rapid cladding oxidation (beginning of Time Frame 2). Timely depressurization (prior to significant cladding oxidation) of the RCS mitigates the tlirtat to the tubes, allows injection of the accumulators and IRWST water, and provides a long term heat sink to establish a controlled, stable state.

Failure to depressurize can rtsult in the failure of the tubes and a loss of containment integrity when oxidation begins.

For saam generator tube rupture (SOTR) initiated sequences, depressurization of the RCS can be used to isolate the faulted steam generator, and re establish core cooling via injection.

De automatic depressurization system (ADS) is required to fully depressurize the RCS to allow the PXS s'/ stems to inject. However, the recovery of passive residual heat removal (FRHR) or injection to the steam generators will provide a substantial heat sink to depressurize the RCS and mitigate the threat to the tubes. 3: CVS cr ': :E;=d :: Se on!Erj ex=d= gmye4: ipa =d:: S RCS =d <C "- '-" *^ '- "- De auxiliary pressurizer sprays are not evaluated for survivability since the inclusion of several other safety related systems which perform the same function provides reasonable assurance I of RCS depressurization in the event of a non LOCA or small LOCA severe acci<ient.

Depressurization of the RCS within Time Frame I is outlined in ERO FR.C 1 which is entered based on high core-exit temperature, ne RCS pressure, core exit temperature and RCS temperature can be used to trionitor the plant response to the RCS depressurization.

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D.6.1.4.2 LOCA Sequences LOCA sequences (other than small LOCA sequences) by definition are depressurized below the secondary system pressure by the initiating event and therefore, are not a threat to steam generator tube integrity upon the onset of rapid oxidation. Depressurization may be required for injection to establish a long term heat sink. Intermediate LOCAs require additional depressurization to allow the injection of RNS or PXS. Medium LOCAs require additional depressurization to allow the injection of PXS. Large LOCAs are fully depressurized by the initiating event.

In LOCA sequences, only the ADS is effective in providing depressurization capability to allow injection to the RCS. Steam generator cooldown and auxiliary pressurizer sprays are not effective.

Depressurization of the RCS is outlined in ERO FR.C 1 which is entered based on high core-exit temperature. De RCS pressure, core-exit temperature and RCS temperature can be used to monitor the plant response to the RCS depressurization.

D.6.1.4.3 Prevent Reactor Vessel Failure Depressurization of the RCS, along with injecting into the containment is an accident management strategy to prevent vessel failure, ne depressurization of the RCS reduces the stresses on the damaged vessel wall faceitating the in vessel retention of core debris, De ADS is used to depressurite the RCS to prevent reactor vessel feilure.

Depressurization of the RCS is outlined in ERO FR.C l. FR.C 1 is entered based on the core exit temperature. De RCS pressure, core exit temperature and RCS temperature can be used to monitor the plant response to the RCS depressurization.

D.6.1.5 Depressuriae Steam Generators The steam generators are depressurized to facilitate low pressure injection into the secondary system and to depressurize the RCS in non LOCA and small LOCA sequencu. Injection to the steam generator must be available to depressurize the secnndary system to prevent creep rupture failure of the tubes, ne steam generator PORV and steam dump valves we used for depressurizing the steam generators.

Depressurization of the steam generators is outlined in ERO FR.H 1 as a means to facilitate low pressure injection into the steam generators, ne steamline pressure and RCS pressure can be used to monitor the plant response.

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4 D. Equipment SunMbtlity Asumment D.6.1.6 Contalament liest Removal Containment heat removal is not explicitly listed as a high level action in the AP600 SAMO Framework, but it is implicit it, the high level action *Depressurire Containment."

Containment heat removalis provided by the passive containment cooling system (PCS). De PCS heat removal through a dry containment shell is sufficient to prevent containment failure; however, water cooling of the shell is needed to establish a controlled, stable state with the containment depressurized. De actuation of PCS water is typically automatic in Time Frame 0.

PCS water is supplied to the extemal surface of the containment shell from the PCS water storage tank or the post 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> water tank. Alternative water sources can be provided via separate connections outside containment, ne containment heat removal can be monitored with the containment pressure and the PCS water flowrate or PCS water storage tank level.

D.6.1,7 Containment Isolation .

Containment isolation is not explicitly listed as a high level action in the AP600 SAMO Framework, but it is implicit as a requirement to protect the fission product barrier.

Containment isolation is provided by an intact containment shell and the containment isolation system (CIS) which closes the isolation valve in lines penetrating the containment shell, ne containment isolation can be monitored by the containment pressure and the CIS valve positions.

D.6. l.8 flydrogen Control Maintaining the containment hydrogen concentration below a globally flammable limit is a requirement for a controlled, stable state, ne containment can withstand the pressuiization from a global deflagration, but potential flame acceleration can produce impulsive loads for which containment integrity is uncertain. While hydrogen is not generated in a significant quantity until Time Frame 2, provisions are provided in the ERGS within Time Frame i to tum on the igniters before hydrogen generation begins so that hydrogen can be bumed as it is pwfuced.

Severe accident hydrogen control in the AP600 is provided by hydrogen igniters. De containment has passive auto-catalytic ircombiners (PARS) as well, but they are not credited t- << vere accidents.

The niters are manually actuated from the control room as the first step in ERG FR.C 1 on high core exit temperature. De intention of this timing is to actuate the igniters prior to the T Westinghouse o vmv.itw wpr skiiton D 10

d D. Equipawat Survivahdity Assessament cladding oxidation (Time Frame 1). De containment hydrogen concentration is monitored prior to actuation so that a globally flammable mixture is not unintentionally ignited.

The plant response to the igniter ac:uation can be monitored by containment hydrogen concenuation using the hydrogen monitors or the post. accident sampling function, which is part of the primary sampling system. De containment pressure response can also be used to observe hydrogen buming.

D.6.1.9 Accident Monitoring Accident monitoring is a post TM1 requirement as outlined in 10 CFR 50.34(f). Aside from the accident management purposes outlined above, monitoring the progression of the accident and radioactive ne',<ases provides input to emergency response and emergency action levels.

Accident monitoring is provided by the in containment monitors for pressure, hydrogen concentration, water levels, and radiation, as well as the post accident sampling system.

D.6.2 Time Frame 2 In Vessel Core Melting and Relocation ,,

Time Frame 2 represents the period of core melting and relocation and the entry into the SAMO. De intact and coolable in vessel core geometry is lost, and relocation of core debris into the lower head is likely. De in vessel hydrogen generation and fission product releases from the fuel matrix occur during this time frame.

D 6.2.1 Injection into the RCS in Time Frame 2, the in vessel core configuration loses its coolable geomeuy and it is likely that at least some of the core debris will migrate to the reactor vessel lower head. If the RCS is depressurized and the reactor vessel is submerged, the core debris will be retained in the reactor vessel. However, injection into the RCS to cover and cool the core debris is required to achieve a controlled, stable state. RCS injection is not required to protect the containment fission product boundary, injection is successful if it is sufficient to quench the sensible heat from the core debris and maintained to remove decay heat.

RCS injection is outlined from SAMO SAO 3 (Reference D 3) and entered from the Diagnostic Flow Chart. Water can be injected into the RCS using i: PXJ, the CVS or the RNS systems, he PXS is not credited in Time Fmme 2 because automatic and manual activation of the system is attempted several times in Time Frame 1 (EROS AE 0 and AFR.C 1), and diverse pumped systems are credited to provide reasonable assurance of RCS injection survivability in this time frame. Post core damage, the actions may be monitored with RCS pressure or temperature or containment pressure.

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a D Equipement Survivability Aueument D.6.2.2 Injrttien into Containtnent ne objective of injection to the containment prior to reactor vessel failure (Time Frame 3) is to cool the extemal surface of the vessel to maintain the core debris in the vessel.

Reasonable assurance of injecting to the containment for in vessel retention is achieved by instructing the operator to drain the IRWST in the EROS within Time Frame 1. After relocation of core debris to the lower head in Time Frame 2, the success of this action becomes uncertain. If the vessel fails, the accident progresses to Time Frame 3. Active operation for injection to containment is completed prior to Time Frame 2.

D.6.2.3 Injection into the Steam Generators in transients and srnall LOCAs, injection into the steam generators is required to be recovered in Time Frame I to be successful. Steam generator tubes or the hot leg nozzles will fail when the cladding oxidation begins at the onset of Time Frame 2. Steam generator injection is not required for LOCAs which depressurize the RCS below the secondary system pressure.

Within Time Frame 2 SAMG, injection can be utilized in unisolated SOTR sequences to maintain the water level on the secondary side for mitigadon of fission product released.

Injecting into the steam generators, along with depressurization of the RCS, is an accident management action to isolate containment or scrub fission products. Failure to inject to the faulted steam generator in Time Frame 2 can lead to continued breech of the containment fission product boundary and large offsite doses.

De main feedwater and startup feedwater pumps are used to inject into a pressurized secondary system. If the secondary system can be depressurized sufficiently, condensate, fire water or service water can also be used to inject into the secondary side.

Injection into the steam generators is covered in the WOO SAMO (Reference D 3)in SAO.l.

He guideline is entered from the Diagnostic Flow Chart based on low steam generator level, either wide range or narrow range. De plant response is monitored with the steam generator level and steamline pressure.

D.6.2.4 Depressurize RCS RCS depressurization is required within Time Frame I for facilitating in. vessel retention of core debris and for successfully prevenung steam generator tube failure in high pressure severe accident sequences. De steam generator tubes or hot leg nozzles will fait due to creep rupture after the onset of rap'd oxidation at the beginning of Time Frame 2. His action facilitates in. vessel retention of core debris in conjunction with injection into the containment to give time to recover pumped injection sources to establish a controlled, stable state.

Reasonable assurance of successful RCS depressurization is provided by inttructing the operator to depressurize the system in the ERGS in Time Frame 1. Active opedon of RCS depressurization is completed prior to Time Frame 2.

D T T WestMgh0tJS8

. . ~ ,, - D.i2

p. r.quipment survinbility Assessment D.6.2.5 Depressurtre Steam Generators Active operation to depressurize the steam generators is used to cooldown the RCS prior to Time Frame 2. After the onset of core melting and relocation, depressurizing steam generators could threaten steam generator tube integrity. Depressurizing the steam generator in Time Frame 2 does not facilitate the establishment of a controlled, stable state.

D.6.2.6 Containment Heat Removal Reasonable assurance of successful containment heat removal is provided since automatic actuation of PCS water occurs in Time Frame 0 and passive air co< : ig of dry shell prevents containment overpressurization, providing time for operator to recover a water source.

Altemate water sources can be provided by connections to the extemal PCS water tank which is outside the containment pressure bouadary and not subjected to the harsh environment.

D.6.2,7 Containment isolation Active operation of containment isolation valves is required in Time Frame 0 or 1 to establish the containment Ossion product bamer. Therefore, only the survivability of the containment pressure boundary, including penetrations, is required to maintain containment isolation after Time Frame 1.

D.6.2.8 Hydrogen Control The operator action to actuate the igniters occurs prior to the hydrogen generation at the onset of Time Frame 2. 'Ihc igniters need to survive and receive power throughout the hydrogen release to maintain the hydrogen concentration below the lower Dammability lirnit during the hydrogen generation in Time Frame 2.

D.6.2.9 Mitigate Fission Product Releases A nonsafety-related containment spray system is provided in AP600 to wash aerosol fission products from the containment atmosphere. The spray system is manually actuated from the SAMO which is entered at the onset of Time Frame 2. Operating the spray inydves opening an air-operated valve inside the containment and actuating valves and a pump outside the containment. Once open, the active operation of the valve inside the containment is l completed.

l D 6.2.910 Accident Monitoring During the initial core melting and relocation, containment hydrogen and radiation monitors are used for core damage assessmeN rd ven0 cation of the hydrogen igniter operation.

Steam generator radiation monitoring is used to determine steam generator tube integrity. In the longer term, the post. accident sampling function can be used to monitor hydrogen and radiation. Containment pressure needs to be monitored throughout Time Frame 2.

Restsion: 11 T Westiligh00$4  % DRAFT D.n . .o _ ,t - s

A D. Equipment Suribability Aswument D.6.3 Time Frame 3. Ex Vessel Core Relocation Time Frame 3 represents the phaw of the accident after vessel failure. The core debris is in the reactor cavity, and the IRWST water is not injected into the containment.

D.6.3.1 injection into the RCS The RCS is failed. Injection to the RCS is no longer needed in Time Frame 3.

D.6.3.2 Injection into Containment

• Reasonable assurance of suf0cient water coverage to the ex. vessel debris bed is passively provided by the containment design to drain water from the RCS, CMTs, and accumulators to the lower containment. Water condensing on the PCS shell is retumed to the reactor cavity after Alling the IRWST and a small volume in the refueling canal to the overnow. Without draining the IRWST water to the cavity, the CMT accumulator and RCS water provides sufficient water retum to the cavity to maintain water coverage over the ex vessel debris bed.

D.6.3.3 Injection into the Steam Generators De RCS is failed. Injection into the steam generators is no longer needed in Time Frame 3.

Injection to the steam generator for SOTR fission product scrubbing is not required to maintain the water level as the water cannot drain against the containment backpressure.

D 6.3.4 Depressurire RCS De RCS is depressurized by the vessel failure in Time Frame 3.

D.6.3.5 Depressurire Steam Generators The RCS is failed. Steam generator depressurization is not needed in Time Frame 3.

D.6.3.6 Containment Heat Removal Active operation of PCS water is completed prior to Time Frame 3.

l D.6.3.7 Containment Isolation and Venting Continued operation of the containment shell as a pressure boundary is needed to maintaic containment isolation in Time Frame 3.

In the event of containment pressuritation above design pressure due to core concrete interaction non.condensable gas generation, the conainment can be vented via the RNS hot leg section. Venting through this pathway relieves to the spent fuel pool. Venting protects l containment isolation by preventing an unconuolled containment failure airbome re; ease

. [ W85tingh00$8

. w w n ii w .pt m oio ws D 14

e D. Equipment Survivability Autument pathway. De vent can be opened and closed as required to maintain pressure in the containment below service 1.evel C. Containment venting does not prevent or mitigate containment basemat failure due to core concrete interaction.

D.6.3.8 Combustible Gas Control The hydrogen igniters are used to control combustible gases. Active operation of igniters continues to control the release of combustible gases from the degradation of concrete in the reactor cavity.

8 D.6.3.9 Mitigate Fission Product Releases ne nonsafety related sprays are actuated in Time Frame 2. He operation of the nonsafety containment spray pump continues, possibly into Time Frame 3. until the water from the source tank is depleted.

l D.6.3.910 Accident Monitoring Containment pressure and the post accident sampling function are sufficient .o monitor th6 accident in the long term. -

D.6.4 Summary of Equipment and Instrumentation De equipment and instrumentation used in achieving a controlled, stable state following a severe accident, and the time it operates are summarized in Tables D.6 2 through D.6-4.

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C D. Equipment Sunivability Auessment Table D61 AP600 tilGli LEVEL ACTIONS RELATIVE TO ACCIDENT htANAGEMENT GOALS (taken from Table $.I. reference D 1)

Goal Element liigh Level Action Controlled, stable core water inventory in RCS + inject into RCS

• depressurire RCS water inventory in containment + inject into containment heat transfer to SO: e inject into RCS

• inject into SGs e depressurire SGs heat transfer to containment + inject into RCS

• inject into containment

+ depressurire RCS Controlled, stable heat transfer from containment

• depressurire containment containment

• vent containment isolation of containment + n,xt into SGs

+ depressunse RCS hydrogen prevention / control + burn hydrogen

+ pressurire containment

• depressurire RCS

• inject into containment S vent containment CCI prevention + inject into containment ilPMB prevention + inject into containment

+ depressurire RCS creep rupture prever. tion e depressurire RCS

• inject into SGs containment vacuum prevention + pressurire containment Terminate fission product isolation of containment e inject into SGs release + depressurire RCS reduce fission product inventory + inject into containment

+ depressurire RCS reduce fission product driving force + depressunte containment g) hg W8Silfigh00$8 o'pr Wu.Il%yd opf it>lll(N7 D.16

Table D.6-2 E M

EQUIPMENT AND INSTRUMENTATION OPERATION FIUOR TO END OF TIME FRAME I - CORE UNCOVERY AND h

S Acties E,*._%

HMWP lastruementatses Purpose Comassend

}I

=F m

1g inject into RCS + PXS

• cae cut t/c's - restore core cooling + injection umst ohen be j CVS RCS pressure

• - recovered so be successful na J

+ RNS - RCS RTDs severe accalent g

- CVS flow 4 q

• RNS flow >

i Inject into + PXS recire - corew:xit s/c's - prevent vessel falure - manual cavity D. g actica Contmament - SFS injection to - coatmament waser in ERG FILC- scred when refuchag cavity level CET/C) 1200*F g

• IRWST waser level Inject into SGs

• High Pressure . SG WR waser level -

naabhsh heat sink - injection source must chen be

-MFW - sacamliac pressure - make SGs ava !able .-J o be successful in

- SFW to depresswue RCS severe accmicat

[ + tow Presswe + prevent creep rupture w . condcasase

- inre waser

- service water Depressurue RCS

• ADS - RCS pressure + facihtase injection to - ADS chen automatac

- PRHR HX - core-cut t/c's RCS

- via SGs - RCS RTDs + long-term heat

- '.= P= S c _, transfer path N + pevent creep rupsure - RCS depresswuanon required

- contarament insegnty pior to ciakhng oxidanon to prevent creep rupture g . esolate break in - uses intact SG or PRHR SGTx C_

. prevent vessel failure + requires injection to

, containment to be successful a

u= '

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fga Table D.6-2 (Cost.)

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EQUIPMENT AND INSTRUMENTATION OPERATION PRIOR TO END OF11ME FRAME I - CORE UNCOVERY AND HEATUP Acties

• j . '. - : -- lastrunwesassee Purpose Ceemund 3 Depressurue SGs - SG FORY - sacandiae pressere - facalatase injere== to - reqawes injection into SGs so

• Secans duanp - RCS pressure SGs prevent creep repare

[

g - depressurue RCS Containment Heat - PCS waner - coatmamcat - cousamuncat insegrwy - PCS waser ofica ammomauc Removal - external fare pressure - alleviase weser - PCS feowrase environmensal

- PCS tark level chalicage to t g... - . - :

- long-serm heat transfer path Contanamcat - CIS - CIS valve posation - costmasneet insegnty - CIS ofica automatic I mle == - contaamment shell - contanament - manual actace in ERG E-0 0 .

• pressaare f l Penetrataoes Comeroi Hydrogen - ignisers - contmames - contnament insegrity - manual iganer action in ERG bydrogen mcenors FR.C-I casered when CEDC

- contmement > 1200*F premare

'N

• SG radiation - accident -- _p - requred by 10 CFR 50340 E Accident

% Monitoring

• contanuneet - emergency resposue p pressure - emergency action

- costanment levels hydrogen E

- containment waser $2 level l

lk ,

- contmament raa - [

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Table D.6-3 9 Y EQUIPMENT AND INSTRUMENTATION OPERATION DURING TIME FRAME 2 -

5' c

$ IN-VESSEL CORE MELTING AND RELOCATION

}

h Action lapt inso RCS a-MGE Espeepenest lastrissme=s=s==e

• RCS pressure Pierpose

• cool can detms Coenament

• RCS agectson needed so coul so-vessel

=

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• CVS

• contasarracna pressure detms for reasonaNe assermsw of 1

. l

• RN3

• CVS flew comuuued. saaNe stase  ;

g

• RNS flow E In p t Inso

• acnic operanuamiEJ na Time 4 33 Contaenment Frarric I >

it Inscx2 anno SGs

• High Pressure

• SG WR water level

• tsoame centsament m

• also reqmres RC3 depressamauon for ll -MFW SGTR -=

g (e

- SFW

• Low Pressure

• scrub fisason products ,

- Condensme

- Fue Waser .

- Service Waer O

• acnic cperason compiesed sa Tune L Depressurue RCS

• Frame 1 Depressurue SGs

• acave operasson compiesed sa Temic Frame I Contannmens Hem

• acave operanon compiesed sa Time Removal Frame I Contamenent Isolanon

• contanment shell - containment pressure

• contamment meegrity

• CIS acave op:rapon complesed sa Time

• penetrations Frame I l

Control Hydrogen

• ognosers

• contasnment hydrogen - a=s-anic:. meegrwy

• acave operaoun concaues in Tome Frame monstors 2

• post-accidens sampimg

• monnors caly requered snessally so venfy fisncteue hydrogen egwer opersson g

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Table D4-3 (Cont.) gl '"i:!

l 1)7-[ EQUIPMENT AND INSTRUMENT ^ TION OPERATION DURING TIME FRAME 2 -

IN-VESSEL CORE MELTING AND RELOCATION g

L U Action F,* _ ;-

jaetrumsteGatsee Purpose CW s

% Control Fisuos spray pwnp gray flowrase scrub aerosols - manual acDon wahra SAMG

~

g Product P,elemes - spray valve - containsneat pressere j Accsdent contaamment pressure mW - acuve operanon coeuaves sa Time

- Moeisoring - post-accident snanagement Frame 2 saanpling fasecuon - esnergency response

- aux tMg r=&=r-m - escrgency accoa nonitors levels

- SG radiauon rnonitors 8

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. '3 y EQUIPMENT AND INSTRUMENTATION OPERATKM DURING TL%8E FRAME 3 - g EX-VESSEL CORE RELOCATION g a

Action lastrumenesesse Pterpose Cw h f

^

3

, Inject iac RCS

- eos needed in Tune Raine 3 Jv Inject seio - injecteos of CMTs and j Consensnent aci.-- '~s in Time Rame I y q provides reasonable assurance of g

waner coverag- to ex-vessel core debris F.

Inject into SGs - not needed in Tune Ranie 3 Depressurtze RCS

- not needed in Time hame 3 O Depressurize SGs

- not needed in Tune Rame 3 u

=

Containment Heat

- acuve operation completed in Removal Tsme Rame I Coniaemment - contmament -

wasmament - comamamen sneegrity - active operassue of CIS -,M isolation shcII pressure in Tanne Frasne I

[

+ penetrations

. RNS hot leg

• coatmasnent vent - meanual acuan within SAMG suction MOVs fg active operation coetmees in

- contaenment insegray -

Control Hydrogen - ignisers - post-accident sampling functsua Time Rame 3 C.

c k

'R W

- C 5"

.k a 5, ?-

o m- 1 k

• O llc 3g0

1. E-Table Dh-4 (Coat)

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='s EQUIPMENT AND INSTRUMENTATION OPERATION DURLNG TIME FRAME 3 -

EX-VESSEL CORE RELOCATION

-(. _

{ Action Y , ', 1-^ Instruseestatsee Ptsepose Ceemament b Control Fnsion - spray penp - spray flowrase - scrub fassion products - active operauco cuounues y Product Ren ew Accident - containment pressere - accalent management - active operation contmues in Monstering - post-accalent - emergency response Time Frame 3 sa M *2 uncuon f - emergency action

- m.a Wg. twiima levels tus -,2s

- 50 <adiation 8DOdtilDr5 -

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D. F.quipment Survivability Assessment l D.7 Severs Accident Environments l D.7.1 Radiation Environment . Severe Accident The radiation exposure inside the containment for a severe accident is conservatively est%ated by considering the dot.c in the middle of the AP600 containment with no credit i e the shielding provided by internal stmetures.

Sources are based on the emergency safeguards system core thermal power rating and the following analytical assumptions:

. Po wer Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1,972 MWt

. Fraction of total core inventory released to the containment atmosphere:

Noble Oases (Xe, Kr) . . . . . . . . . . . . . . . . ....................... 1.0 H alog en s (1, B r) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0.7 5 Alkali Metals (Cs, Rb) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0.7 5 Tellurium Group (Te, Sb, Se) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0.305, B arium, Strontium (Ba, S r) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0.104 Noble Metals (Ru, Rh, Pd, Mo, Tc, Co) , . . . . . . . . . . . . . . . . . . . . . . . . . 0.005 Lanthanides (La, Zr, Nd, Eu, Nb, Pr, Sm. Y. Cm, Am) . . . . . . . . . . . . . . 0.0051 Cenum Group (Ce, Pu, Np) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0.0051 De radionuclide groups and elemental release fractions listed above are consistent with the accident source term information presented in NUREO 1465, " Accident Source Tenns for Light Water Nuclear Power Plants Final Repon," with the exception of the early in. vessel release fractions for barium and strontium, lanthanides, and cerium groups. For banum and strontium, the value of 0.004 is used in pixe of the NUREG 1465 value of 0.02. For the lanthanide and cerium groups, the release fraction of 0.0001 is considered in place of the NUREO 1465 values of 0.0002 for lanthanides and 0.0005 for the cerium group. Rese exceptions are based on the recommendations of the Depanment of Energy's Advanced Reactor Severe Accident Program (ARSAP) in support of tne ALWR program (Ref. D-4).

De timing of the releases are based on NUREG 1465 assumptions a well as AP600 specific activity release projections. The release scenario assumed in the calculations is desenbed below.

An initial release of activiy from the gaps of a small number of failed fuel rods at 30 seconds into the accident is considered. De instantaneous release of 0,15 percent of the core inventory of the volatile species (defined as noble gases, halogens, and alkali metals) is assumed. At 50 minutes after the accident, an additional 2.85 percent of the core activity inventory is assumed to be instantaneously released from the gaps of failed fuel rods and is added to the previously released inventory associated with 0.15 percent of the gap activity.

T Westinghouse b D.23 o sprawv.Itw pt ib-t:0197

4 D. Equipment Survivability Assessment Dus, the total release of volatile species at 50 minutes after the accident is 3 percent of the total core inventory. 1 At 30 minutes following the instantaneous gap activity releases, that is,80 minutes into the i accident, an additional 2 percent of the core inventory is added to the inventory that exists based on the previous gap activity releases. At this point,5 percent of the total core inventory of volatile species has been assumed to be released.

Over the next 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, releases associated with an early in vessel release period are assumed to occur, that is, flom 1.33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> (or 80 minutes) to 2.63 hours7.291667e-4 days <br />0.0175 hours <br />1.041667e-4 weeks <br />2.39715e-5 months <br /> into the accident. His source term is a time varying release in which the release rate is assumed to be constant during the duration t!me, consistent with the assumptions in NUREO 1465. Additional releases during the early in vessel release period include 95 percent of the noble gases 35 percent of the halogens, and 25 percent of the alkall metals, as well as the fractions of the tellurium group, barium and strontium, noble metals, lanthanides, and cerium group as listed above, ne doration of the ex. vessel release is two hours and the late in vessel release is ten hours.

Rese releases occur simultaneously after the early in vessel release. De additional releases include 35 percent of the halogens,45 percent of the alkali metals, over 25 percent of the tellurium group, ten percent barium and strontium and fractions of the noble metals ~,

lanthanides and cesium group, consistent with the assumptions of NUREO 1465.

De resulting instantaneous gamma and beta dose rates are provided in Figures DA04D.71 and DAO-3D.7 2, respectively.

l D.7.2 Thermal Hydraulic Environments l D.7.12.1 ":::f!:;; m::: A:d& t Ent:r::tControlled Hydrogen Combustion The bounding severe accident environments for each of the equipment survivability time frames defined in Section D.4 are provided in this section, nese bounding environments for the reactor coolant system and containment are used in the assessments ascribed in Section D.8.

De MAAP4 computer code (version 4.0.2) was used to support the quanti 0 cation of the equi 9 ment survivability time frames and the bounding environment within each time frame.

Two basic sequences and five sensitiv.y cases were quantified to establish the bounding environments including hydrogen combustion in the cont inment. Each sequence input data were adjusted to assure that a 100% fuel clad metal wate6 reaction occurred so that the required bounding hydrogen source was considered.

De two base sequences were a large (2.2 ft 3) hot leg break into a steam generator compartment and a 4 inch direct vessel injection (DVI) line break in a valve vault room. For each of these I.OCA sequences, four sensitivity cases were run to determine the effects of cavity Gooding, core concrete interaction, igniters (local burn versus global burn) and jet o \pr2Wy.llupp d wpr It>l20197 D 24

O. Equipment Survivability Assemenwat buming of the heated hydrogen-rich RCS gas discharge. A total of ten sequences were quantified. De designcor and description for each of the ten sequences are summarized in Table D.71.

ne key event timing for each of the sequences is summarized in Table D.7 2. Rese key events in the severe accident progression directly relate to the equipment survivability time frames. Time Frame 1 is the interval between core uncoven and a core exit gas temperature exceeding 2000*F (1367 K). Time Framc 2 is the interval _between the core exit gas temperature exceeding 1367 K and either the end cf core material relocation into the lower

- head or vessel failure. Time frame 3 is the interval between vessel failure and the end of the sequence, ne MAAP4 results provide the botmding containment environment associated with the combustion of hydrogen resulting from the equivalent of 100% oxidation of the active fuel cladding where: 1) igniters are functioning (local buming scenario),2) igniters were artificially defeated (global buming scenario), and 3) jet buming and igniters were defeated (global buming scenario). To calculate more severe bounding containment environments, the cavity flooding was defeated in some sequences resulting in ex vessel hydrogen generation due to core-concrete interaction.

ne results of 4 inch DVI line break sequences are very similar to the hot leg large LOCA results because the ADS 4th stage valves are opened in both sequences. The RCS response for these low pressure sequences is very similar. De peak temperature calculated in the upper plenum gas was about 2780*F (1800 K). Since these sequences are low pressure sequences with the ADS 4th stage valves open, the gas temperature in the pressurizer stayed below the nominal temperature (665'F,625 K) for most of the transies in all of these sequences. The gas temperatures in both steam generators stayed below 566'F (570 K) for all of these sequences because water was present in both steam generator secondary sides.

l Figures D-A44D.7 3 through DA4-6D.7 8 show gas temperatures in the containment compartments, the containment pressure and the RPV temperature. Since this sequence has cavity flooding, resulting in no vessel failure and no core concrete interaction, all hydrogen bums occuned before 11,000 secor.ds. However, the hydrogen burned was not due to igniter induced bums. Some of the hydrogen coming out fmm the primary system through the ADS 4th stage valves were bumed as they came out because the primary system gas temperature was higher than the jet bum temperature (1448'F,1060 K). To see the effect of the jet bum, the jet bum model v.as tumed off in the A3BE GJ and APLHL-GJ sequences. For cases without the jet bum, it was observed that a large amount of hydrogen was bumed in the upper compartnant and much less hydrogen was bumed in the steam generator compartments. De containment gas temperatures after 11,000 seconds reached stable conditions because of the availability of PCS.

In general, results with and without igniters were very similar because of the jet bum of the gas flow coming out from the primary system. For cases without cavity flooding, more Revision: 11 ENE.

3 M figh00$$ mL DRAET D 25 W .II W *Pr 4 120lM

a l

l l

D. Equhpawnt Survivabuity Assessawat i

hydrogen was generated due to the core concrete interaction such that late hydrogen bums were observed.

i For the hot leg LOCA sequences with the cavity Gooding available, the water level in the  ;

containment eventually reached the hot leg break elevation and the whole core became i submerged by water in the later uansient. The reverse water flow through the break did not occur in the DVI line break sequences because of the h!;-): Goor elevation of the valve vault ne families of curves for the other base case (large LOCA) and all the mere sity cases are provided in Figures D444D.7 9 through D..! ED.7 62.

l l D 7.2.2 Global Hydrogen Combustion This section presents the containment environmental response to a scenario in which the hydrogen generated from 100 percent cladding reaction is allow-i to mix in the cor.tainment and then is globally burned. The case presented here is the same as case FRF1 from the AP600 hydrogen combustion analysis (Chapter 41), except that, after the cladding reaction is completed and the hydrogen is mixed in the containment, a bum is initiated that produces a pressure and temperature spike in the containment. De hydrogen mass genemted by the sequence, and the pressure and temperature resp.:.nse in the containment are presented in Figures D.7-63 through D.7 69.

A 'I U@M o wawv.t Iw=pt ib I:0197 D 26

a D. Equipment Servivability Assessment Table D.71 SEQUENCE DESIGNATOR Hot leg LOCA vith a break area of 2.2 ft2 with no reflood: 2 ADS Stage 13 and 4 ADS Stage 4, I accumulator. I CMT, and PCS available.

- APLHL Large hot leg with cavity flooding, similar to the A3BE, except the break location and the area. The break location is in S/G compartment I, rather than valve vault.

. APLHL N APLHL + no cavity Gooding + no ex vessel cooling

- APLHL-G APLHL + no igniters

- APLHL-GN APLHL + no igniters + no cavity flooding + no ex vessel cooling

. APLHL-GJ APLHL + no igniters + no jet burn 4 inch DVI line break with no redood and no PRHR: 2 ADS Stage 13 and 4 ADS Stage 4. I accumulator, 1 CMT, and PCS available.

- A3BE AP600 3BE sequence with cavity flooding, ex vessel cooling, igniters, and jet burnig

- A3BE N A3BE + no cavity flooding + no ex vessel cooling

- A3BE G A3BE + no igniters

- A3BE GN A3BE + no igniters + no cavity flooding + no ex vessel cooling

- A3BE GJ A3BE + no igniters + no jet burn W6dflgh00$4 D-27 o W av.tl W *#1611 N

• C 30 1:

_j5a

t Table D.7-2 (h{r

SUMMARY

OF MAAP4 ANALYSES: EQUIPMENT SURVIVABILITY TIME FRAMES j _

xe, -, -

TW 44 mets DVI Line Bremit see=1s Het Lag tarse LOCA 3 A3SE A3SE-N A3SFA A3SE-CN A3SFAJ AFLitL ArtJEL-N AFLHEA AFLHLCN AFtJ8tAJ Ctat Oamisaca (%) 100 41 9536 100 4 800 43 100t47 100 42 100 4 800 43 100 49 100 44

[

~g Tune d Case 2M7 2M6 2767 2M6 2M7 2006 20E2 2006 20E2 2006 thew::ry (s)

Tenne d Core Emm Gas 4262 42 % 4262 42 % 4262 34 % 3453 34 % 3453 34 %

Tesap. > IM7 K Tune d Inshal CDec 5662 5692 5700 508 5700 4949 4924 4949 4925 4949 Masenal Reb ==*a=

isso tower Head Time d Vennel Famiwe N/A 20589 N/A IM52 N/A N/A 18:20 NA 19119 NA Tune core Masenal 16800 20100 20700 16300 18500 - 14000 38500 14u00 15000 14000 Relocasion isso tower Head Ends O

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1 D. Equipment Survivability Assessment D.8 Assessment of Equipment Survivability Since severe accidents are very low probability events, the NRC recommends in SECY 93 087, that equipment desired to be available following a severe accident need not be subject to the qualification requirements of 10CFR50.49, the quality assurance requirements of 10CFR50 Appendix B, or the redundancy / diversity requirements of 10CFR50 Appendix A.

It is satisfactory to provide reasonable assurance that the designated equipment will operate following a severe accident by comparing the AP600 severe accident environments to design basis event / severe accident testing or by design practices.

D.8.1 Approach to Equipment Survivability The approach to survivability is by equipment type, equipment locatic,n, survival time i required, and the use of design basis event qualification requirements and severe environment experimental data.

D.8.1.1 Equipment Type The various types of equipment needed to perform the ::::::d:g activities discussed aboye are transmitters, thermocouples,' resistance temperature detectors (RTDs), hydrogen and I radiation monitors, ::!:::!d valves, pumps, valve limit switches, containment penetration assemblies, igniters, and cables.

D.8.1.2 Equipment Location Some of the in containment equipment, i.e. transmitters, have been deliberately located to avoid the most severe calculated environments. Other equipment is located outside containment, ne performance of all the equipment was judged based on the most severe postulated event for that location.

D.8.1.3 Time Duration Required I ne .:: Sed:; requirements have inn defined for each time frame, so the equipment evaluation only discusses performance during these periods. Time Frame I ends between 3453 seconds (0.96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />) and 4262 seconds (1.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) depending on the event. Time Frame 2 ends between 14000 seconds (3.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) and 38500 seconds (10.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />). A limited amount of equipment has been designated for the long term (Time Frame 3) and these parameters can be monitored outside containment.

D.8.1.4 Severe Environment Experiments The primary sout:e for performance expectations of similar equipment in severe accident environments is EPRI NP-4354,"Large Scale Hydrogen Burn Equipment Experiments". His information is supplemented by NUREG/CR 5334, " Severe Accident Testing of Electrical Penetration As:cmblies." nese programs tested equipment types that had previously been w [ W65tiflgh0US8 o evn.irapp.d 2pr ib i:oie7 D-98

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l D. Equipement Survivability ha=== nt qualified for design basis event environments! conditiore. The temperature in the chamber for the first program was in ?he 700 800*F range 'or ten to twenty minutes during the continuous hydrogen injection tests. Although the conditions at the equipment would be somewhat less severe, the chamber conditions c.ivelop all of the longer duration profiles indicated for the AP600 evenu. De equipment in this program was also exposed to significant hydrogen bum spikes that are also postulated for the APtiOO. De same equipment was exposed to and survived several events, both pre mixed and continuous hydrogen injection which provides confidence in its ability to survive a postulated severe accident. He second program tested containment penetrations to high temperature,s for long durations. The Westinghouse penetration was tested under severe accident conditions simulated with steam up to 400"F and 75 psia for ten days. De results indicated that the electrical performance < '

the penetration would not lead to degraded equipment performance for the first four days, ne mechanical performance did not degrade (no leaks) during the entire test.

D.8.2 Equ'pment Located in Containment ne exposure to elevated temperatures as a direct result of the postulated severe accident or as a result of hydrogen buming is the primary panmeter of interest. Pressure environments will not exceed the design basis eve;.. conditions for which the equipment has been qualified.

Radiation environments also will not exceed the design basis event conditions throughout Time Frames 1 & 2.

D.8.2.1 Differential Pressure and Pressure Transmitters The functions defined for severe accident management that utilize in-containment transmitters are IRWST water level, reactor coolant system pressure, steam generator wide range water level and containment pressure. Most of these transmitters that provide this information are located in the valve rooms where the environment is limited to short duration temperature transients. Dese traiaients exceed ambient design basis temperature conditions but should not impact the transmitter perfcrmance since the intemal transmitter temperature will not increase significantly alwve that experienced during design basis testing. EPRI NP-4354 documents transmitter performance during several temperature transients with acceptable results, ne IRWST water level transmitters are located in the lower compartment and are only required during Time Frame 1. De environment during Time Frame I will not exceed the design basis qualification parameters of the transmitters. Reactor system pressure and steam generator wide range water level are required through the second time fame, ne only long term application is the containment pressure transmitter which may eventually be impacted by the severe accident radiation dose, but containment pressure could also be measured outside containment if necessary.

D.8.2.2 Thermocouples ne functions defined for severe accident management that utilize thennocouples are core exit temperature and containment water level. De core exit temperature (locted in the upper plenum) is only required during Time Frame I and the containment wate' level (located in T Westigh0088 f D-99 W ilWPC # 15tMl"

1 D. Equipment Survivability Assessment j <

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the steam generator I compartment and the cavity) is required through Time Frame 2. The i temperatures to which the thermocouples are exposed during the defined time frames do not exceed the thermocouple deSQn.

D.8.2.3 Resistance Temperature Detectors (RTDs)

Both hot and cold leg temperatures are defined as parameters for severe accident management in Time Frame 1. RTDs are utilized for these measurements and will perform until their temperature range is exceeded. The hot leg RTDs could fail as the temperature increases well above the design crnditions of the RTDs but the cold leg R7Ds should perform throughout Time Frame 1.

D.8.2.4 Ilydrogen Monitors Containment hydrogen is defined as a parameter to be monitored throughout the severe accident scenarios. Early in the accident, the hydrogen will be monitored by a design basis event qualified device that operates on the basis of catalytic oxidation of hydrogen on a heated element. De hydrogen monitors are located in the main containment area. De design limip of this device may be exceeded ~

after the first few hours (2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) of some of the postulated xcidents and performance will be uncertain. If the device fails, hydrogen concentration will be determined through the post. accident sampling function.

D.8.2.5 Radiation Monitors Containment radiation is defined as a parameter to be monitored throughout the severe accident scenarios. The containment radiation monitors are located in the main containment area. Early in the accident, the design basis event qualified containment radiation monitor will provide the necessary information until the environment exceeds the design limits of the monitor (2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> for some events). 11 the device fails, containment radiation will be determined through the post. accident sampling function.

D.8.2.6 Solenoid Valve I Aeees: :: 6: = d::=: = ;== =: f== 6: p : =d& ,: = p!!:; f= d= !: d= ugh I  : ='nid :p=rd vdv: !::=M t- $:!::=-:: p===: 5 = .::::=: 'hi:5 1:

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[l) _ Qualified solenoid valves are used to vent air operated valves (AOVs)'to perform the function -

I / required. in Time Frame 1, the core makeup. tank AOVs located in the valve vault provide 1 - a path for RCS injection the PRHR AOVs located in steam generator compartment i provide I a path for RCS depressurization and the containment is isolated by AOVs located in the lower l= compartment and the valve vault. The environment to which these solenoid valves may be i exposed in Time' Frame I is not signif' antly different than _the design basis events to 'which I  ; the devices are qualified. In Time Frame 2, the RCS boundary AOV located in steam 1 -- generator compartment 1 is required for CVS injection into' the RCS and the containment

, i_l spray AOV located in the lower compartment is required for control of fission product release.

In additioa throughout Time Frame 3, access to the containment environment from the post- 1 i l accident sampling function is through solenoid valves located in the lower compartment.

'l! 'During Time Frames 2 and 3, these valves may be exposed to transient conditions due to 17 hydrogen bums that exceed design basis event qualification. Solenoid valves in_ an energized condition were included in the hydrogen burn experiments (EPRI NP 4354) and survived )

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.I. accident radiation dose will not exceed typical design basis qualification for these valves.

-l l' D.8.2.7 ' Motor Oper5ted Valves ,

-1 I_ Motor operated valves (MOVs) are utilized in several applications during the severe accident I scenarios.- MOVs in-the accumulator and core makeup tank path are normally open and I remain open. In Time Frame 1 the PXS recirculation MOVs located in the valve vault are l_ ' required for injection into the containment, MOVs for the first three stages of ADS located

! -in steam generator compartment I are required for RCS depressurization and the containment

. I is isolated by MOVs located ir, *.he lower compartment and the valve vault. The environment I I- to which these MOVs may be exposed in Time Frame I is not significantly different than the

i. L design basis events to which Wey are qualified. In Time Frame 2, the charging and injection

. 'I MOV located in the lower compartment provides a path from the CVS for RCS injection and I an RNS MOV located in the valve vault provides a path from the IRWST for RCS injection.

< l In addition, throughout 'fime Frame 3, containment venting to the spent fuel pool is available I through RNS hot leg suction line MOVs located in the valve vault. During Time Frames 2 I and 3, these valves may be exposed to transient conditions due to hydrogen bums that exceed

< :I design basis event qual 4fication. MOVs were included in the hydrogen burn experiments C 1: (EPRI NP 4354) and rurvived many transients. Shielding provided by the location of the l~ ~ valve ensures that the severe accident radiation dose will not exceed typical design basis I qualification for these valves.

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D.8.2.8 Squib Valves Squib valves are only required in Time Frame I when the severe accident environment is not L

significantly different than the design basis environment for which these valves are qualified.

li IRWST and PXS recirculation squib valves located in the valve vault are used for injection L - into the RCS and containment, respectively. For RCS depressurization, the fourth stage ADS b squib valves are located in steam generator compartments I and 2.

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Ti D. Equipment Survivability Assessment I D.8.2.79 Limit Switches Limit switches are required to monitor the position of containment isolation valves that could lead directly to an atmospheric release. Dese isolation valves will close early in the transient, so verification is only required during Tiaie Framc 1. The limit switches are located in the lower compartment and the environment in this time frame will not exceed the design basis i event qualification of the limit switches.

I I D.8.2.810 Hydrogen Ignitrrs j De hydrogen igniters are distributed throughout the containment and are designed to perform in environments similar to those postulated for severe accidents. He igniters transformers are located outside containment. The successful results of glow plug testing through several hydtcgen bums is documented in EPRI NP-4354 and provides confidence in the performance of these devices.

I D.8.2.911 Electrical Containment Penetration Assemblies ne electrical containment penetrations are located in the lower compartment and are requ to perform both electrically and mechanically throughout the severe accident. He hydrogen burn equipment experiments documented by EPRI NP-4354 included a Westinghouse penetration qualified for nuclear plants. Electrical testing on the penetration cables after all the pre mixed and continuous injection tests concluded that most (39 of 52) of the cables passed the electrical tests while submerged in water. Dese tests consisted of ac (at rated voltage) and dc (at three times rated voltage) withstand tests and insulation resistance tests at 500 volts. The Westinghouse penetration was also tested under simulated severe accident conditions at 400*F and 75 psin for about 10 days (NUREG/CR 5334), he results indicated that some degradation in instrumentation connected to the penetration may occur in four days under these severe conditions. He lower compartment may experL ,ce short temperature transients above 400*F but stable temperatures are significantly less, so it is expected that the electrical performance would be maintained throughout the event. De only long term measurement utilizing these penetrations is containment pressure and this can easily be measured outside containment if necessary. Here was no degradation of mechanical i performance (maintaining the seal) in either test program. He mechanical performance of I penetrations is also discussed in Chapter 42, 1 D.8.2.1012 Cabhe The hydrogen bum equipment experiments documented by EPRI NP-4354 included twenty-four different cable types qualified for nuclear plants. Electrical testing on these cables after all the pre mixed and continuous injection tests concluded that all (fifty two samples) of the cables passed the electrical tests while submerged. These tests consisted of ac (at rated voltage) and dc (at three times rared voltage) withstand tests and insulation resistance tests at 500 volts. Due to the exposure to many events, some cable samples had extensive damage in the form of charring. cracidng and bulging of the outer jackets and still performed Rension: 11 DRAFT T Westiflgt10Use o Wawv.iiw .pt iw 20197 D-102

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satisfactorily. De cables tested are representative of cables specified for the AP600 and will only be exposed to short single temperature transients in their respective locations. Proper performance can be expected. The only long term measurement utilizing cables is containment pressure, which can be measured outside containment if necessary.

D.8J Equipment Located Outside Containment i Other functions defined for severe accident management are menneredperformed outside containment and are not subjected to the harsh environment of the event. These include the I steamline radiation monitor and transmitters for monitoring steamline pressure; the passive I containment cooling system flow and tank level; and-the post accident sampling function; the i CVS pumps and flow measurement; the RNS pumps and flow measurement; SFS MOV for I injection to the IRWST; all methods of injection into the steam generatots, i.e., MFW pumps I and valves, SFW pumps and valves and condensate, fire water and service water pumps and i valves; steam generator PORVs and steam dump valves for depressurization; PCS valves and I fire water pumps and valves for containment heat removal; containment isolation valves; I auxiliary building radiation monitor; MOV and manual valve from RNS hot leg suction lines I to the spent fuel pool; and containment spray pump, valves and flow measurement. ,

D.9 Conclusions of Equipment Survivability Assessment ne equipment defined for severe accident management was reviawed for performance during the environments postulated for these events. Survivability of the equipment was evaluated based on design basis event qualification testing, severe accident testing, and the survival time I required following the initiation of the severe accident. It is concluded that the equipment, I all of which is qualified for design basis events, has a high probability of surviving portulated severe accident events and pert'orming satisfactorily for the time required.

AP600 provides reasonable assurance that equioment, both electrical and mechanical, used to mitigate the consequences of severe accidents and achieve a controlled, stable state can i perform over the time span for which they are needed. Containment stn.ctural integrity is I discussed in Chapter 42.

D.10 References D-l Framework for AP600 Severe Accident Afanagement Guidance WCAP 13914 Revision I, November 1996.

D-2 AP600 Emergency Response Guidelines, Revision 3, May 1997.

D3 Westinghouse Owneri Group Severe Accident Management Guidance, June 1994.

D-4 Letter from B. A. McIntyre, Westinghouse, to T. Quay, NRC, "AP600 Loss of Coolant Accident Source Term Model," NSD-NRC-96-4675, April 1,1996.

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