ML20141K278
| ML20141K278 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 04/18/1997 |
| From: | Joseph Sebrosky NRC (Affiliation Not Assigned) |
| To: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 9705280425 | |
| Download: ML20141K278 (9) | |
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April 18,1997 Mr. Nicholas J. Liparulo, Manager Nuclear Safety and Regulatory Analysis Nuclear and Advanced Technology Division Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, PA 15230
SUBJECT:
FOLLOWON QUESTIONS REGARDING THE AP600 INSPECTIONS, TESTS, ANALYSES, AND ACCEPTANCE CRITERIA (ITAAC)
Dear Mr. Liparulo:
As a result of its review of the June 1992,-application for design certifica-tion of the AP600, the staff:has determined that it needs additional informa-tion.
Specifically, the enclosure'to this letter contains requests for_
i additional information concerning the AP600 ITAAC resulting from a review done by the Containment Systems and Severe Accident Branch.
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 i
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 followon ques-i 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 accordance 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) 415-1132.
Sincerely, l
original signed by:
Joseph M. Sebrosky, Project Manager Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No.52-003
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Enclosure:
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i DOCUMENT NAME: A:SCSB ITC.RAI (6F AP600 DISK)
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OFFICE PM:PDST:DRPM SPA:PDST:DRPMd D:PDST:DRPM l l
NAME JMSebrosky:sg) & JNWilsoffV TRQuay W DATE 04/[(,/97
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Letter to Mr. Nicholas J.' Lioarulo. Dated: Aoril 18. 1997
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- Docket File.
PUBLIC PDST R/F TMartin
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MSlosson TQuay TKenyon WHuffman JSebrosky DJackson-JMoore, 0-15 B18 WDean, 0-17 G21~
JNWilson ACRS (11)
JLyons, 0-8 E23 ACubbage 0-8 E23 Alevin, 0-8 E23 1
JBongarra, 0-9 H15 l
TCheng, 0-7 H15 GThomas, 0-8 E23 MChiramal,0-8 H3 DThatcher, 0-7 E4 j
HWalker, 0-8 D1 JLyons, 0-8 D1 REmch, 0-10 D4 JBongarra, 0-9 HIS JPeralta, 0-9 Al JKudrick, 0-8 H7 MSnodderly, 0-8 H7 EThrom, 0-8 H7 BPalla, 0-8 H7 Blong, 0-8 H7 i
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I Mr. Nicholas J. Liparulo Docket No.52-003 Westinghouse Electric Corporation AP600 cc: Mr. B. A. McIntyre Mr. Ronald Simard, Director Advanced Plant Safety & Licensing Advanced Reactor Programs Westinghouse Electric Corporation Nuclear Energy Institute Energy Systems Business Unit 1776 Eye Street, N.W.
P.O. Box 355 Suite 300 Pittsburgh, PA 15230 Washington, DC 20006-3706 Ms. Cindy L. Haag Ms. Lynn Connor Advanced Plant Safety & Licensing Doc-Search Associates
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Westinghouse Electric Corporation Post Office Box 34 Energy Systems Business Unit Cabin John, MD 20818 4
Box 355 i
Pittsburgh, PA 15230 Mr. James E. Quinn, Projects Manager i
LMR and SBWR Programs Mr. M. D. Beaumont GE Nuclear Energy Nuclear and Advanced Technology Division 175 Curtner Avenue, M/C 165 Westinghouse Electric Corporation San Jose, CA 95125 One Montrose Metro 11921 Rockville Pike Mr. Robert H. Buchholz Suite 350 GE Nuclear Energy Rockville, MD 20852 175 Curtner Avenue, MC-781 San Jose, CA 95125 Mr. Sterling Franks U.S. Department of Energy Barton Z. Cowan, Esq.
NE-50 Eckert Seamans Cherin & Mellott 19901 Germantown Road 600 Grant Street 42nd Floor Germantown, MD 20874 Pittsburgh, PA 15219 Mr. S. M. Modro Mr. Ed Rodwell, Manager Nuclear Systems Analysis Technologies PWR Design Certification Lockheed Idaho Technologies Company Electric Power Research Institute Post Office Box 1625 3412 Hillview Avenue Idaho Falls, ID 83415 Palo Alto, CA 94303 Mr. Frank A. Ross Mr. Charles Thompson, Nuclear Engineer U.S. Department of Energy, NE-42 AP600 Certification Office of LWR Safety and Technology NE-50 19901 Germantown Road 19901 Germantown Road Germantown, MD 20874 Germantown, MD 20874 t
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Followon Questions on the AP600 Certified Desian Materials (CDM)
Containment Systems and Severe Accident Branch Comments 640.55 The ITAAC in Section 2.2.1 do not include stroketimes for the containment isolation valves.
i 640.56 The following comments relate to Section 2.3.9, " Containment Hydro-gen Control System," of the CDM which involve the hydrogen recombi-l nation subsystem and the hydrogen ignition subsystem. The hydrogen recombination subsystem provides hydrogen control during and follow-ing a design basis LOCA while the hydrogen ignition subsystem provides hydrogen control during and following a degraded core or core melt scenario.
For the hydrogen recombination subsystem, the applicant has speci-fled (a) Design Commitments, (b) Inspection, Tests, Analyses, and
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(c) Acceptance Criteria in Table 2.3.9-2 of the CDM Document.
The HRS is provided to meet the requirements of GDC 41, " Containment i
Atmosphere Cleanup," 10 CFR 50.44, " Standards for Combustible Gas Control Systems in Light-Water-Cooled Power Reactors." Westinghouse references RG 1.7, " Control of Combustible Gas Concentrations in Containment Following A Loss-Of-Coolant Accident," in the SSAR as the methodology used for implementing these regulations. The ITAAC fail to address these requirements and Westinghouse's commitment to the methodology in RG 1.7.
a)
Section 6.2.4.1.1, " Containment Mixing," of the SSAR is to i
provide an analysis which shows that excessive stratification of combustible gases will not occur within the containment or within a containment subcompartment. Verification of the analysis that supports the design commitment to provide a system to mix the combustible gases within containment has 1
been omitted from the ITAAC.
b)
Design Commitment No. 3.a of Table 2.3.9-2 fails to verify conformance with design criteria, such as NUREG-0737, "Clari-fication of TMI Action Plan Requirements," Item II.F.1,
c)
Design Commitment No. 3.b of Table 2.3.9-2 verifies the existence of a report that establishes the depletion rate for a single full-size PAR.
There is no link between this accep-tance criteria and the installed PARS.
d)
Either the criteria used to locate the PARS inside containment or a description of their specific location inside containment should be provided in the ITAAC.
Because the criteria used to l
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judgement, the staff recommends the use of a detailed descrip-tion or figure to verify appropriate location of the PARS.
e)
The ITAAC should verify the existence of a report that con-cludes that the installed PARS are qualified for a harsh environment and can withstand the environmental conditions that would exist before, during, and following a design basis accident without loss of safety function for the time required to perform the safety function. The report should also address the potential of the fission products that make up the post-accident radiation environment to be catalytic poisons.
f)
For the hydrogen ignition subsystem (HIS), the applicant has specified (a) Design Comitments, (b) Inspection, Tests, Analyses, and (c) Acceptance Criteria in Table 2.3.9-2 of the CDM Document. The HIS is provided to safely accommodate hydrogen generated by the equivalent of a 100 percent fuel-clad metal water reaction as required by 10 CFR
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50.34(f)(2)(ix). The system also ensures that uniformly distributed hydrogen concentrations in the containment do not i
exceed 10 percent (by volume). This is accomplished by initiating a deflagration at the lower level of hydrogen flammability.
The ITAAC fail to verify several important design features of the HIS as described in Section 6.2.4.2.3 of the SSAR.
The igniters have been divided into two power groups.
Power to each group will be normally provided by offsite power, however should offsite power be unavailable, then each of the power groups is powered by one of the onsite non-essential diesels and finally should the diesels fail to provide power then approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of igniter operation is supported by the non-Class IE batteries for each group.
Assignment of igniters to each group is based on providing coverage for each compart-ment or area by at least one igniter from each group.
The igniter assembly is designed to maintain the surface tempera-ture within a range of 1600 to 1700*F. These design features are essential in establishing the HIS's ability to initiate a deflagration at the lower level of hydrogen flammability and should be verified by the ITAAC.
g)
Either the criteria used to locate the igniters inside con-tainment or a descriptio') of their specific location inside containment should be provided in the ITAAC.
Because the criteria used to locate the recombiners was subjective and based on engineering judgement, the staff recommends the use of a detailed description or figure to verify appropriate location of the recombiners.
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~ 640.57 The following comments relate to Section 2.2.2, " Passive Containment Cooling System," of the CDM.
a)
A report should be prepared as part of Design Commitment 6.a) to provide documentation that the integrated flow from the three PCS flow phases,.in combination with the inventory verification under Design Commitment 6.e), assures that the PCCWST can provide cooling water for the required 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period. The minimum flow requirements of Design Commit-ment 6.a) are not sufficient. Documentation of the measured flow rates can also be used to detemine degraded flow capa-bility over the life of the plant.
b)
The acceptance criteria for Design Commitment 6.b) needs to include a measurement of the surface area coverage from the PCS water at the upper spring line for each of the three phases of the PCS flow. The minimum coverage fractions need to be verified and consistent with the water distribution test, for example at least 90 percent coverage for the initial phase.
In addition it needs to be confirmed that the side wall watet coverage is consistent with the water distribution test, both in minimum area coverage and uniformity around the circumference. Documentation of the measured coverage frac-tions and uniformity of the flow can also be used to determine degraded surface conditions over the life of the plant.
c)
Recent design changes to the PCS to address post 72-hour actions in response to the staff requirements memorandum of January 15, 1997, on SECY-96-128, " Policy and Key Technical Issues Pertaining to the Westinghouse AP600 Standardized Passive Reactor Design," have not been incorporated into the ITAAC. The most recent description of the design changes is provided in Westinghouse letter NSD-NRC-97-5024, "AP600 Design Changes to Address Post 72-hour Actions," B. A. McIntyre to T. R. Quay, dated March 14, 1997. New design features include increased inventory in the PCCWST, the addition of an on-grade PCS auxiliary water storage tank, and two recirculation pumps which provide the required makeup flow to the PCCWST from the auxiliary tank for the post 72-hour period (for up to~seven days).
In addition, the PCCWST now also provides makeup to the spent fuel pool (SFP) and the interface between the PCS j
and the SFP systems have not been included in the ITAAC.
i The Design Description, Figure 2.2.2-1 and Table 2.2.2-1 need to be updated to include the new post 72-hour design features and the SFP interface.
d)
The acceptance criterion for Design Commitment 6.e) needs to be updated to the new PCCWST inventory, and as appropriate CDM 2.3.7, " Spent Fuel Pool Cooling System," needs to be updated to include the PCCWST interface requirements.
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Design Commitment 6.f), concerning long-term makeup to the PCCWST needs to be modified, and if necessary additional sub-sections added, to address the new post 72-hour design features.
For example, demonstration that each recirculation pump can deliver the required flow rate to the PCCWST, that the on-grade PCS auxiliary water storage tank is seismically qualified and can withstand wind and tornado loadings, instru-mentation is available to measure water level in the tank, the pump can be supplied from the on-site diesel generator, and verification of the minimum volume of the auxiliary storage tank.
f)
The design basis performance of the PCS requires that the containment be coated with an-inorganic paint on the exterior surface to enhance surface wetability and PCS water area coverage, and on the interior surface to promote development of the condensation film. Adequate PCS water area coverage on the exterior surface is also based on a system of weirs (referenced to here as water collection troughs, Tag No.
PCS-MT-04) which collect and uniformly redistribute the PCS water from the water distribution bucket to the upper spring line of the containment shell. The ITAAC does not address either the inorganic paint or the uniform distribution of the PCS water over the exterior shell surface.
The Design Description needs to be updated to include the inorganic paint as part of the PCS.
Requirements for surface preparation, for example requirements of the Steel Structures Painting Council, and application of the paint to its required thickness, based on paint manufacture's requirements and consistent with the Westinghouse-PCS test program, need to be included.
1 g)
The design basis performance of the PCS is also based on adequate heat removal from the containment atmosphere by
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internal, structural heat sinks. The design basis analyses of the PCS performance is based on a maximum steel jacket-to-concrete air gap thickness which accounts for shrinkage of the i
concrete over the life of the plant. An increased air gap i
thickness will reduce the effective heat transfer and result in an increase in the containment pressure response following a design basis accident. To assure that this maximum air gap thickness is not exceeded over the life of the plant, the concrete composition (for example aggregate size and moisture content), steel T-pin length and initial pour need to be controlled and verified. The ITAAC does not address the minimum air gap thickness.
Suitable requirements concerning the concrete composition, steel T-pin design and initial pour need to be included in the Design Description.
Figure 2.2.2-1 and Table 2.2.2-1 should also be modified as appropriate.
e,
-h)
Schematic Figure 2.2.2-1 does not indicate the presence of the combination flow restricting and flow measuring orifice on each PCS water delivery line, as shown in Figure 6.2.2-1,
" Passive Containment Cooling System Piping and Instrumentation Diagram," of the SAR, Revision 11.
Installation of the proper orifice on each line is essential to the PCS performance.
Figure 2.2.2-1 and Table 2.2.2-1 should be updated to include these orifices.
640.58 The following comments concern Severe Accident Mitigation Features.
a)
Draft SRP Section 14.3.11 has the reviewer ensure that appro-priate treatment of severe accident design features and containment design features are included in Tier 1.
The supporting information regarding the detailed design and analyses should remain in Tier 2.
For many of the design features, it may be impractical to test their functionality because of the absence of simulated severe accident condi-tions. Consequently, the existence of the feature on a figure, subject to a basic configuration walkdown, may be considered sufficient Tier 1 treatment.
Design features essential to maintaining containment integrity and assuring a low conditional containment failure probability (CCFP) in severe accidents should be selected for treatment in Tier 1.
For AP600, these systems would include the reactor cavity flooding system, the hydrogen igniter system, and the ability to manually depressurize the RCS following core damage, since these features are critical to maintaining a low CCFP as shown in Chapter 50 of the PRA. Westinghouse should provide cross references in the appropriate sections of Tier 2 to show how the design features and SSCs found important from PRA, exter-nal event analyses, shutdown risk study, and severe accident '
analyses are verified by the ITAAC. Westinghouse has not adequately addressed severe accident design features in the CDM or provided cross references to show how the important insights or assumptions from the PRA are verified by the ITAAC.
b)
Design criteria for severe accident mitigative features are contained in SECY-93-087, " Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor Designs." The Commission approved these design criteria in an SRM dated July 21, 1993. At a minimum, the key systems and features provided in the AP600 design to address the criteria described in the hydrogen control, core debris coolability, high-pressure core melt ejection, containment performance, dedicated containment vent penetration, and equipment survivability sections of SECY-93-087 should be provided in the ITAAC.
c)
One of the more important assumptions in the PRA is the high probability of maintaining reactor vessel integrity during a
a.
. core melt scenario. Heat is removed from the molten core debris through boiling on the outside of the flooded reactor vessel. This phenomena is often referred to as in-vessel retention in the PRA.
In order to credit the in-vessel retention approach,two design objectives must be met.
- First, the cavity flooding system must cover the lower reactor vessel prior to relocation of the core.
Important design criteria for the cavity flooding system are addressed in Section 2.2.3,
" Passive Core Cooling System," of the CDM. Second, the reactor insulation system must allow the ingress of water and not interfere with the boiling process.
Some of the important criteria to meet this design objective are:
flow paths and clearances, ball and cage check valve design, steam vent damper design, ability to sustain differential pressure loads given in Section 39 of the PRA, provisions to prevent plugging by debris. These important criteria associated with the reactor insulation design should be incorporated into the ITAAC.
d)
Another important assumption is the containment's ability to meet Service Level C.
Important design criteria associated with the containment shell, equipment hatr.h, electrical per9tntions and mechanical bellows should be verified in the ITAAC.
640.59 The containment recirculation screens and IRWST strainers are designed in conformance with RG 1.82 and to address the technical concerns raised in Generic Safety Issue A-43, " Containment Emergency Sump' Performance." The staff provided further clarification of these concerns in Bulletins 96-03 and 95-02.
The staff disagrees with Westinghouse's assessment that these bulletins only apply to boiling water reactor designs and are not applicable to the AP600 strainer design. Westinghouse needs to document in the SSAR what actions will be taken to address the conditions described in these bulleti1s on the IRWST strainer design.
Design commitments result-ing fro..i the review of RG 1.82, GSI A-43, and the bulletins should be incorporated into the ITAAC.