ML20198N649

Forwards Open Items Associated W/Chapter 19 of AP600 SER

ML20198N649
Person / Time
Site:	05200003
Issue date:	11/19/1997
From:	Joseph Sebrosky NRC (Affiliation Not Assigned)
To:	Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
References
NUDOCS 9801210157
Download: ML20198N649 (11)

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4 Y November. 19,-1997- . s -

4 7.f ,i'NMr? Nicholas Ji Liparulo, Manager -

, Nuclear Safety- and Regulatory Analysis:

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SUBJECT:

.OPEN ITEMS ASSOCIATED WITH CHAPTER-19 0F THE AP600 SAFETY EVALUATION) '

j ; , d. 4 y REPORT (SER)-

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Dear Mr. Liparulo- -

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The Containment Systeins'and Severe Accident Branch has provided an SER fori JChapter 19.--The SER provided was for the Level 2 and 3 portion of the probabilistic risk assessment. However, the_ input to these sections contained 3 T..some open. items. ,These open items have been extracted from the SER and can- -

been found in Enclosure 1 to.this letter, i

.You have requested that portions of the information submitted in the

' June 1992, application for design certification be exempt from mandatory

' 'public disclosure. While the staff has not completed its review of your request in accordance with the requirements of 10 CFR 2.790, that portion of the submitted information is being withheld from public disclosure pending the staff's final determination. The staff concludes that these follow on ques-tions do not contain those portions of the information for which exemption is sought. However, the staff will withhold this letter from public disclosure for.30. calendar days from the date of this letter to allow Westinghouse the opportunity to verify the staff's conclusions. If, after that time, you do not request that all or portions of the-information in the enclosures be withheld from public disclosure in accordar.ce with 10 CFR 2.790, this letter will be placed in the Nuclear Regulatory Commission Public Document Room.

-If you have any questions regarding this matter, you may contact me at (301) a15-1132.

Sincerely, original signed by:

Joseph M. Sebrosky, Project ManagerL Standardization Project Directorate Division oi Reactor Program Management Office of Nuclear Reactor Regulation Docket No.52-003 ,

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r Mr. Nicholas J. Liparulo Docket No.52-003 Westinghouse Electric Corporation AP600 cc: Mr. B. A. McIntyre Ms. Cindy L. Haag Advanced Plant Safety & Licensing Advanced Plcnt Safety & Licensing Westinghouse Electric Corporation Westinghouse Electric Corporation Energy Systems Business Unit Energy Systems Business Unit P.O. Box 355 Box 355 Pittsburgh, PA 15230 Pittsburgh, PA 15230 Enclosure to be distributed to the following addressees after the result of the proprietary evaluation is received from Westinghouse:

Mr. Russ Bell Ms. Lynn Connor Senior Project Manager, Programs DOC-Search Associates Nuclear Energy Institute Post Office Box 34 1776 I Street, NW Cabin John, MD 20818 Suite 300 Washington, DC 20006-3706 Mr. Robert H. Buchholz GE Nuclear Energy Dr. Craig D. Sawyer, Manager 175 Curtner Avenue, MC-781 Advanced Reactor Programs San Jose, CA 95125 GE Nucletr Energy 175 Curtner Avenue, MC-754 Mr. Sterling Franks San Jose, CA 95125 U.S. Department of Energy NE-50 Barton Z. Cowan, Esq. 19901 Germantown Road Eckert Seamans Cherin & Mellott Germantown, MD 20874 600 Grant Street 42nd Floor Pittsburgh, PA 15219 Mr. Charles Thompson, Nuclear Engineer AP600 Certification Mr. Frank A. Ross NE-50 U.S. Department of Energy, NE-42 19901 Germantown Road Office of LWR Safety and Technology Germantown, MD 20874 19901 Germantown Road Germantown, MD 20874 Mr. Ed Rodwell, Manager PWR Design Certification Electric Power Research Institute 3412 Hillview Avenue Palo Alto, CA 94303

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- OPEN ITEM ASSOCIATED WITH CHAPTER 19

. (Level 2 and 3 Portion of the Probabilistic Risk Assessment)-

l 720.435F In the Level 2 analysis, the challenge to the RCS pressure boundary in an ATWS -

- eventtis considered to be successfully mitigated U the primary-side pressure remains below 3200 psig throughout the' pressure . . . . :snt. According to WEC, '

LOFTRAN analyses demonstrate that primary side prissure will not-exceed 3200-psig if: all of the following conditions are met:- 4 of.4 reactor coolant pumps c

automatically trip,1 of 2 core makeup tank lines inject,:and 1 of-2 valves on

- the'PRHR open. However, the LOFTRAN analyses were not submitted. In response to staff review comments, WEC determined that the fault-trees for the'CMT and RCP subsystems used in the Level 2 quantification for ATWS incorrectly

- included operator actions.- Because of the limited time available for operator action'in an ATWS, WEC revised the relevant fault trees to remove the operator actions and requantified the PRA for . 71 dent class 3A. The requantification produced'an increase in the containment failure frequency of 2E-10/y

- (1 percent) for the baseline PRA, and 2E-8/y (4 percent) for the focussed PRA, i which the staff considers insignificant. The staff requested that WEC provide additional information regarding the LOFTRAN analyses, and confirm that the analyses are representative of (or conservatively bound) the thermal hydraulic

- response for all Level I core melt sequences assigned to accident class 3A.

-The-staff is awaiting WEC's response. This is Open Item 19.1.3.2.2-1.

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720i436F: The following was extracted from the SER input titled "Early Containment Failure."

The non-safety grade containment spray system was added to the AP600 design subsequent to Revision 8 of the PRA. As such, its-impact on containment

, response is not reflected in the Level 2 and 3 PRA results. Although the use of sprays is generally considered to be beneficial in terms of reducing

- containment pressure and enhancing fission product removal, the sprays could adversely impact containment response by increasing the likelihood and i magnitude of hydrogen combustion events. Specifically, with the addition of sprays, hydrogen deflagrations or deflagration-to-detonation transition (DDT) could-occur in events:in which combustion-would have otherwise been precluded

. by steam inerting, and the magnitude of the deflagrations and likelihood of DDT could increase as a result. of higher pre-burn hydrogen concentrations.

The staff has requested WEC to assess the impact of the non-safety grade containment sprays on hydrogen combustion modelling and assu:aptions in:the

Level _2 analysis, and to confirm that containment performance (and containment failure frequency) will not be adversely impacted.- The staff is. awaiting =

WEC's response, and will confirm the adequacy of WEC's modelling of early m - containment failure following receipt of this information. This is Open Item 19.1.3.2.2-2.-

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2-720.437F: The following discussion was extracted from the input titled

" Intermediate Containment Failure."

As discussed in question 720.436F, the staff has requested WEC to assess the impact of the non-safety grade containment sprays on hydrogen combustion modelling and assumptions in the Level 2 analysis, and to confirm that containment performance (and containment failure frequency) will not be adversely impacted. The staff is awaiting WEC's response. This was identi-fled as Open item 19.1.3.2.2-2. The staff will confirm the adequacy of WEC's modelling _of intermediate containment failure as part of this open item.

720.438F: The following discussion was extracted from the SER input titled

" Frequency and Conditional Probability of Containment Failure."

The CCFP for AP600 is approximately 10.8 percent for the baseline PRA, and 7.1 percent for the focussed PRA for internal events. These values increase to 15.4 and 7.7 percent in the baseline and focussed PRA if diffusion flames at the IRWST vents are assumed to result in containment failure. Intuitively, CCFP should increase in the focussed PRA since the non-safety hydrogen igniter system is removed from service in the focussed PRA (the battery backup to the igniters and the safety grade passive autocatalytic recombiners are not credited in either the baseline or focussed PRA). However, in the focussed PRA any increases in containment failure frequency due to hydrogen igniter unavailability are more than offset by the introduction of ATWS sequences that are conservatively assumed to result in core damage in the Level 1 analysis, but assessed further in the Level 2 analysis and found to not pose a threat to containment integrity, even without igniters. Specifically, the dominant contributors to core damage frequency in the focussed PRA are ATWS events which are assumed to result in core damage due to inadequate pressure relief and localized failures of hardware / software. However, based on LOFTRAN analyses performed by WEC subsequent to the Level 1 PRA, the Level 2 ATWS analysis assumes that the pressure transient will be successfully mitigated if: (1) reactor coolant pumps automatically trip, (2) core makeup tank lines inject, and (3) PRHR operates. Since these functions are generally available in the majority of the core damage sequences, most ATWS events are success-fully mitigated without challenge to containment. This exemplifies how CCFP can be biased by the results of the Level 1 analysis, and illustrates the need to exercise caution in using CCFP as a measure of containment performance. As discussed in question 720.435F, the staff requested additional information from WEC regarding the LOFTRAN analyses, and is awaiting WEC's response. This was identified as Open Item 19.1.3.2.2-1. The inconsistency between the Level 1 and Level 2 modelling of ATWS sequences will be further assessed as part of this open item.

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720.439F Enclosure 2 contains the staff's insights as a result of the review of the Level 2 PRA.- Enclosure 2 contains additional insights from those contained in Chapter 59 of Westinghouse's PRA. Incorporation of the additional insights that exist in Enclosure 2 to the Westinghouse insights is an open item. In addition, the staff believes that certain insights are so important that they need to be incorporated into the Technical Specifications or into the inspec-tions, tests, analyses, and acceptance criteria (ITAAC). In the cases where the staff believes the disposition of the insight should be Technical Specifi-cations or ITAAC a separate FSER open item number has been assigned.

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. i Staff Insiehts as a Result of the Revise of the AP600 Level 2 PRA l 720.440F Passive Coriainment Coolina System flooding of the PCS annulus due to plugging of the upper annulus drains is the only PRA-postulated mechanism for the failure of PCS cooling. The probability of pluaging is minimited in the design by including two 100 percent drains, and a week' y surveillance of the annulus floor and drains to identify and >

eliminate debris that can potentially plug the drains. WEC har credited this .

surve'11ance in the PRA but has not incorporated it in the Technical specifi- '

cations. This is Open itse 720.440F.

• Qgen Items 720.441F and 720.442Fr Reactor Cavity Floodina System A safety-related reactor cavity flooding system is included in the AP600 t design tr' prevent reactor vessel breach and ex-vessel phenomena in the event of a severe accident. The system is comprised of the followi g design features:

two 6-inch diameter recirculation lines that provide a path for gravity draining the IRWST to the reactor cavity.

- a squib valve and a motor operated valve in each recirculation line, each

, powered from the Class 1E de pwer supply, and actuated from the control room, and 1

a reactor vessel thermal insulation system designed specifically to enhance RPV cooling, as described in FSER Section 19.2.3.3.1.

The IRWST injection squib valves are diverse from the containment recircula-tion squib valves. Diversity between these valves is specified in SSAR Section 6.3.2.2.8.9, but the criteria for confirming that diversity has been achieved is not provided. This needs to be addres:;ed by ITAAC. This is Open Item 720.441F.

The containment recirculation squib valves and isolation MOVs, and containment recirculation screens are includea as risk-significant SSCs within D-RAP. -

Surveillance and maintenance requirements on the related piping and valves are provided in the In-Service Inspection and Yesting Programs.

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The operator action to flood the reactor cavity is provided in Emergency [

Response Guideline FR.C-1, which instructs the operator to flood the reactor cavity.if-injection to the RCS cannot be recovered or containment radiation reaches-levels that indicate fission product releases as determined by a core

. damage assessment guideline.

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2- i Key aspects of the reactor cavity flooding system and the containment layout i need to be confirmed by ITAAC to assure that the reactor cavity will flood and  ;

the ApV will reflood as modelled in the pRA (by gravity draining and by manual  ;

actuation of the cavity flooding system). The ITAAC should inc ude confirma-  !

- tion of internal volumer, elevations and inter-compartment vent and drain  !

pathsofthesubcompartmentscontainInsRCspipingcomponentsandimpacting t reactor cavity flooding and RpV rei'looding. WEC needs to provide this ITAAC.

This is open item 720.44tF.  :

720.40F: RPV Thermal Insulation System The AP600 design includes a reflective reactor vessel insulation system that provides an engineered flow path to allow the ingression of water and venting -

of steam for externally cooling the vessel in the event of a severe accident '

involving core relocation to the lower plenum. Key attributes of the insula-tion system are:

RPV/ insulation panel clearances, water entrance and sti.am exit flow I areas, and loss coefficients based on scale tests in the ULPU facility,  ?

ball-rnd-cage check valves and steam vent dampers at the entrance and '

exit of the insulation boundary that open due to buoyant forces during cavity flood-up, and insulation panels and support members designed to withstand the pressure differential loading due to the IVR boiling phenomena.

No coatings are applied to the outside surface of the reactor vessel which will inhibit the wettability of the surface.

The reactor vessel insulation system should be included as a risk-significant

$5C in the reliability assurance program, and reliability / availability controls and goals should be provided, consistent with maintenance rule guidelines, to assure that operability of the system and moving parts is maintained. ITAAC and availability controls are also needed to assure that l- the RPV insulation system will wrform as designed. WEC needs to provide 1

those commitments and ITAAC. Tais is Open Item 720.443F.

l 2fa.444F: Prottetion of Containment From Diffusion Flames The containment layout prevents the formation of diffusion flames that can  !

challenge the integrity of the containment shell. Specifically: '

the openings from the accumulator rooms and CVS compartments that can l vent hydrogen to the CMT room are either located awa l ment wall:and electrical penetration junction boxes,yorfrom'the contain-are covered by a s secure hatch, and IRWST vents near the containment wall are oriented to direct releases  ;

away from the containment shell.

These provisions need to be confirmed by ITAAC. WEC has not provided this

ITAAC. This-is Open Item 720.444F.

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3 Operation of ADS stage 4 provides a vent path for the severe accident hydrogen to the steam generator compartments, bypassing the IRWST, and mitigating the conditions required to produce a diffusion flame near the containment wall.

Note: The following are insights that Westinghouse needs to address in PRA Chapter 59. The insight concerning the " Reactor Cavity Design for Core Concrete Interactions" and the " Containment Vent" contain open items that were previously sent to Westinghouse. These open items are documented in the respective insights.

Containment Isolation System Containment isolation valves in lines that represent risk-significant release paths are controlled by DAS in addition to PMS to further limit offsite releases following core melt accidents. These lines are: containment air filter supply and exhaust, RCDT out, and normal containment sump. The containment isolation valves controlled by DAS are included as risk-signifi-cant SSCs within D-PAP. The operability of DAS actuation of these isolation valves is addressed by sfiort term availability controls for DAS.

Reactor Cavity Desian for Direct Containmerit Heatina The reactor cavity and RPV arrangement provides no direct flow path for the transport of particulated molten debris from the reactor cavity to the 'oper containment regions.

Reactor Cavity Desian for Ex-Vessel Fuel-Coolant Interactions The design can withstand a best-estimate ex-vessel steam explosion without loss of containment integrity.

Reactor Cavity Desian for Core Concrete Interactions --

The reactor cavity design incorporates features that protect against basemat melt-through in the event of RPV failure. The cavity design includes:

8 a minimum floor area of 48 m available for spreading of the molten core debris, layout, elevations, and flow areas of the reactor cavity and RCDT subcompartments and interconnecting ventilation duct consistent with Figure B-3 of Appendix B of the PRA and the supporting ANL analysis, a minimum thickness of concrete above the embedded containmer.1 liner of 0.85 m, provisions te prevent core debris from passing into the sump via inter-connecting pipelines embedded in the concrete floor and/or sump curb, ar.d a sump curb of sufficient height and width to prevent molten core debris frcm overflowing or ablating through the curb.

WEC still needs to confirm these items as part of Open Item 19.2.3.3.3-1 (0 pen Item 720.418F).

A specific type of concrete is not specified for use in the baserat.

Hydroaen Ioniter System The AP600 design includes a hydrogen igniter system to limit the concentration of hydrogen in the containment during severe accidents. The features of the system are:

66 glow plug igniters distributed throughout the containment powered from the non-safety-related onsite ac power systere, but also capable of being powered by offsite ac power, onsite non-essential diesel generators, or non Class lE batteries yta de-to-ac inverters.

manually actuated from the control room when core exit temperature exceeds 1200F, as the first step in ERG FR.C-1 to ensure that the igniter act'"-tion occurs prior to rapid cladding oxidation.

The igniter system is non-safety-related but is subject to investment protec-tion short-term availability controls.

The AP600 design also includes four passive autocatalytic recombiners (PARS) strategically located within the containment. The PARS are provided primarily to cope with hydrogen production during design basis accidents, but are expected to function to reduce combustible gas concentrations during severe accidents.

Non-safety Containment Sorav A nnn-safety grade containment spray system is included in the AP600 design with the capability to supply water to the containment spray header from an external source in the event of a severe accident. Loss of ac power does not i contribute significantly to the core damage frequency: therefore, non-safety-related containment spray does not need to be ac independent. The spray system comprises the followirig design features:

-two containment spray ring-headers equipped with a total of G6 spray nortles and providing approximately 80 percent containment coverage.

a 6-inch diameter supply pipe conne: ting the spray ring-headers to the fire protection system header inside containment, containing or,e normally closed, air-nperated valve with remote actuation from the control room (V701), and one normally open, manual valve (V700),

6-inch diameter piping connecting the fire header inside containment to the fire main header outside containment, and capable of being supplied from both the diesel-driven fire pump and the motor-driven fire pump.

The detailed design and location of all associated valves and connections will take into account expected radiation levels and shielding riquirements for any required local operator actions.-

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The COL applicant will develop and ikplement guidance and procedures for use of the non-safety containment spray system as part of the COL Action item regarding accident management program.

Containment Vent The following will be completed after W submits design description. This is identified as an Open Item in FSER Section 19.2.5 (0 pen Item 720.421F).

1 In the event of a severe accident that results in gradual containment pressur-ization, the AP600 containment can be vented via the line to prevent over-pressure failure.

Fission product releases from the ___ line are routed to the stack.

Valves in the line are qualified to operate at containment pressures corresponding to Service Level C.

The line is capable of withstanding the pressures associated with vent actuiTTon at Service C.

Detailed procedures for use of the containment vent system will be developed by the CO. applicant, as part of the COL Action Item regarding accident management.

This section contingent on WEC's response to staff request concerning provid-ingaventpursuantto10CFR50.34(f)(3)(iv).

Accident Manaaement The COL will develop and implement severe accident management guidance aad procedures using the framework provided 'n WCAP-13914, Revision 2.

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