ML20148H537
| ML20148H537 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 06/05/1997 |
| From: | Slosson M NRC (Affiliation Not Assigned) |
| To: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| SECY-96-128-C, NUDOCS 9706100332 | |
| Download: ML20148H537 (11) | |
Text
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pun q a k UNITED STATES
- 5 g NUCLEAR REGULATORY COMMISSION
- WASHINGTON, D.C. 20066 4001
- * * * * ,o June 5,1997
, Mr. Nicholas J. Liparulo, Manager Nuclear Safety and Regulatory Analysis
- - Nuclear and Advanced Technology Division l Westinghouse Electric Corporation g /'3 s
P.O. Box 355 i Pittsburgh, PA 15230
SUBJECT:
AP600 DESIGN CHANGES TO ADDRESS POST-72 HOUR ACTIONS
Dear Mr. Liparulo:
In SECY-96-128, dated June 12, 1996, the Nuclear Regulatory Commission (NRC) staff recommended for Commission approval that the AP600 be capable of sus-taining all design basis events with onsite equipment and supplies for the long term. After 7 days, replenishment of consumables such as diesel fuel oil l from offsite suppliers can be credited. The equipment required after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> i need not be in automatic standby response mode, but must be readily available for connection and be protected from natural phenomena including seismic events. In the staff requirements memorandum on SECY-96-128, dated January 15, 1997, the Commission approved the staff's position on post-72 hour actions. On February 4, 1997, Westinghouse presented the staff with an overview of proposed changes to the AP600 desigr. to comply with the NRC policy on post-72 hour actions. The staff provided consnents on the proposed AP600 design changes in a February 18, 1997, letter to Westinghouse. Westinghouse letter NSD-NRC-97-5024, dated March 14, 1997, provided responses to the staff's comments on the proposed post-72 hour design changes.
1 Enclosed is the staff's assessment of each Westinghouse response. Based on l the response information, the staff has several concerns. Specifically: l (a) SECY-96-128 states that th'e post-72 hour equipment should be protected I from natural phenomena including seismic events. Westinghouse has l stated that the post-72 hour structures, systems, and components will !
remain functional following an safe-shutdown earthquake and will !
withstand winds up to 145 mph. On April 22, 1997, Westinghouse made a !
presentation at an AP600 senior management meeting concerning how the i post-72 hour equipment is being protected from natural phenomena. In i addition, Westinghouse restated its position on the protection of l post-72 hour equipment from natural phenomena in a letter to the NRC dated April 30, 1997. The staff is still as:;essing the degree of i protection from natural phenomena that is necessary for post-72 hour l equipment. 1 PFoS,i l .
NHC FRE CENIB1 COPY g0090 9706100332 970605 '
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l Mr.' Nicholas J. Liparulo June,5, 19'97 l
l l (b) Indirectly re'iated to the post-72 hour design changes, the staff has focused its attention on the acceptability of the main control room i
emergency habihbility system (VES). It is-the staff's position that the VES flow. rates and maximum carbon' dioxide levels should be consis-tent with the standards used by the American Society of Heating, Refrigeration and Air Conditioning. Engineers. The VES should be capable of supporting the maximum identified number of control room occupants (currently 11) assuming a failure of a. single train.
1 If you have any questions on this matter,'please contact Bill Huffman at
..(301) 415-1141.
Sincerely, original signed by: I s
Marylee M. Slosson, Acting Director Division of Reactor Program Manage a Office of Nuclear Reactor Regulatio?
Docket No'52-003
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NAME WCHuffman:setee x BSherg f,Holahan TQuay A W MS1osson l\V DATE 05f7/o7, 05/t2497 W/')/97 Ob/'5 /97 05/6/97 0FFICIAL LECORD COPY i.
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l Mr. Nicholas J. Liparulo Docket No.52-003 l- Westinghouse Electric Corporation AP600 l
l 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 l Mr. Cindy L. Haag Ms. Lynn Connor i Advanced Plant Safety & Licensing Doc-Search Associates '
Westinghouse Electric Corporation Post Office Box 34 l Energy Systems Business Unit Cabin John, MD 20818 L
Box 355 l Pittsburgh, PA 15230 Mr. James E. Quinn, Projects Manager 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 i Suite 350 GE Nuclear Energy l Rockville, MD 20852 175 Curtner Avenue, MC-781 i San Jose, CA 95125 Mr. Sterling Franks l r ;
U.S. Department of 1A$egy Barton Z. Cowan, Esq-l '
NE-50 Eckert Seamans Cherin & Mellott '
19901 Germantown Road 600 Grant Street 42nd Floor :
l Germantown, MD 20874 Pittsburgh, PA 15219 i
l 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 3t22 Hillview Avenue Idaho Falls, ID 83415 Paio Alto, CA 94303 Mr. Frank A. Ross Mr. Charles Thompson, Nuclear Engineer U.S. Department of Energy, NE-42 AP600 Certification Office of LHR Safety and Technology NE-50 19901 Germantown Road 19901 Germantown Road Germantown, MD 20874 Germantown, MD 20874 l
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u Staff Assesssent of Westinghouse Responses to NRC Comments on AP600 Design Changes
)
to Address Post-72 Hour Action Policy J 4
Reference:
Westinghouse Letter NSD-NRC-97-5024 dated March 14, 1997 Assessment of Responses to NRC Concerns (a) Since Westinghouse is not taking credit for the spent fuel pool heating, ventilation, and air conditioning (HVAC) system, it has performed analyses to demonstrate that offsite dose rates from spent fuel pool 4
boiling are a small fraction of 10 CFR Part 20 dose limits. The staff cannot reach a conclusion about the acceptability of this approach until 4
review of the supporting dose rate analyses submitted by Westinghouse in
' a letter dated April 7,1997, is completed. -In addition, the staff needs additional information related to potential harmful environsent from spent fuel pool boiling discussed in item (b) below. I (b) Post-72 Hour Ventilation After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> into a design basis accident with the loss of offsite and onsite power, Westinghouse had previously proposed to provide natural I circulation ventilation of the main control room (MCR) by opening the control room doors. Westinghouse now proposes to use small ancillary fans to assist in the cooling and ventilation of the MCR. The fans would ;
be powered by the post-72 hour ancillary diesels and would draw outside air into the main control room. The staff is considering the revised Westinghouse proposal, however, additional information is required in ;
order to assess the proposal. Therefore, Westinghouse should update the standard safety analysis report (SSAR) by adding a subsection to i Section 9.4 on the " Ancillary Ventilation System" with the appropriate i text, figures and tables including a revision to Table 3.2-3.
Toxic Releases SSAR Section 6.4.7 should be updated to state that the COL Applicants are responsible for compliance with Regulatory Guides (RGs) 1.78, "Assump-tions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Pnstulated Hazardous Chemical Release" and 1.95, "Protec- '
tion of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release." ,
i Enclosure
i 2
Potential Harmful Environment Generated by Boiling of the Spent Fuel Pool (e.g., steam, radiological elements)
Westinghouse states that the analysis of spent fuel pool boiling shows that spent fuel pool steam and vapor is vented from the fuel bandling area to the outside environment via a blowout panel. This vent location is more than 150 feet from the MCR. Westinghouse should discuss the process by which boiling of the spent fuel pool causes the blowout panel to open including the pressure differential required to activate the blowout panel, leaktigtness of the spent fuel pool area, and provide the calculations which substantiates these conclusions. Westinghouse should provide a detailed evaluation, including calculations of the environmen-tal effects on safety related (and non-safety.related equipment which may impact safety-related equipment) due to pressure buildup in the spent fuel pool area as a result of spent fuel pool boiling prior to the lifting of the blowout panel. Westinghouse should also provide addi-tional basis for its conclusion that the MCR envelope and surrounding safety related equipment will not be affected by the spent fuel pool boiling.
(c) The staff's Comment (c) requested Westinghouse to justify its conclusion i that the global effect of the design changes to the seismic response of J the nuclear island is negligible by appropriate calculations and :
evaluation. The response to this comment states that Westinghouse has initiated aapropriate calculations and evaluation to support its l judgement t1at the global effect of the design change is negligible and I the detail and schedule of the resolution tasks are described in the response to Comments (11), (12), and (13). The NRC staff does not believe that it is feasible to demonstrate the acceptability of the PCS tank and the shield building roof responses without a seismic reas-sessment of the entire nuclear island structures. When this seismic reassessment is performed, Westinghouse should use the seismic model with properly accounted live load masses in the analysis.
Assessments of Responses to NRC Comments
.(1) ... response to the staff's concern with the 18 gpm flow rate, Westing-house has now added a new, non-safety related, 400,000 gallon grade level water storage tank as part of the PCS for the post-72 hour period. The passive containment cooling system (PCS) recirculation pumps will be used to provided make<-up to the passive containment cooling water storage tank (PCCWST) and keep the minimum PCS flow at the 50 gpm rate from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to seven days. The PCCWST inventory will still be increased from 400,000 gallons to 550,000 gallons, to accommo-date spent fuel pool make-up.
The staff has still not seen eq detailed description of the new water storage tank. Consequently, a inal conclusion on the acceptability of the design cannot be reached untii the oesign changes and revised MG0THIC analysis are submitted. The staff recognizes that the design
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of the grade level storage tank is contingent on the staff's position l regarding the appropriate protection of the post-77 hour equipment from j natural phenomena. !
(2) In response to the staff's concern with the pressure increase above the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> criterion, Westinghouse indicated that the analysis provided :
during the February 4, 1997, meeting had errors. A revised analysis !
with corrected input will be performed. The Westinghouse acceptance l criterion of maintaining the containment pressure below 50 percent l of design pressure is retained. The staff finds this response :
satisfactory.
(3) Westinghouse states that there will be no changes to the ERGS to ;
reflect the post-72 hour actions and design changes. Post-72 hour scenarios will be dealt with in the site's accident management plan.
The staff finds this response satisfactory, but requests that Westing-house identify the need to address post-72 hour actions as part of the COL action item Westinghouse will be providing related to accident management.
(4) Westinghouse stated that analyses will be performed to assess the operator doses in-the main control room from spent fuel pool boiling and justify the iodine partition factor in the spent fuel pool. This analysis was submitted by Westinghouse in a letter dated April 7,1997.
The staff is currently reviewing this submittal.
(5) Westinghouse states that the offsite and main control room doses will be assessed over a 30 day time period. In addition, Westinghouse noted thtt the main control room doses will be evaluated in accordance with GDC 19 and that offsite doses due to spent fuel pool boiling will be 3 i assessed in accordance with 10 CFR Part 20. The staff finds this !
response satisfactory.
(6) Concern about Unfiltered in-leakage into the Main Control Room The AP600 design features such as a vestibule style entrance preventing
- contaminated air from entering the MCR envelope as a result of egress and ingress, and maintaining the MCR envelope at positive pressure, with respect to surrounding areas, are not uncommon. This design is !
quite prevalent at operating reactors and those reactors have compara-tively much larger amounts (i.e., hundreds of cfm) of unfiltered i
.in-leakage _as has been determined by a tracer gas testing method. l Non-safety-related ductwork and associated equipment routed through the '
MCR envelope, such as the AP600 VBS design, can provide pathways for unfiltered in-leakage of contaminated air from outside the MCR envelope !
- and also helps to provide a false in,dication of pressurization of the MCD envelope. Pressurization testing measurements do not assure that the MCR envelope is at e uniform positive pressure with respect to i i adjacent areas because pockets in the MCR envelope may be at a negative i pressure with respect to the adjacent areas and thus providing pathways i
- for relatively larger amounts of additional unfiltered in-leakage than l
o t l
! 4 assumed in the dose calculations. This assessment is valid at most ,
operating. plants and it has required reassessment of dose calculations
! for postulated design basis accident conditions. i Therefore, Westinghouse should validate the VES design. The NRC staff has determine that an acceptable method of validation is testing in ;
accordance with ASTM E741, " Standard Test Method for determining Air Leakage Rate by Tracer Dilution." Westinghouse should also update SSAR ,
Sections 6.4, 9.4.1, and the Technical Specifications Section to reflect these changes.
(7) Main Control Emergency ventilation System (VES) Flow Rate, NCR Envelope Occupancy and Carbon Dioxide Level The AP600 ventilation design uses references (for example, those published in the 1954,1961, and 1969 time frame) which do not reflect i current staff positions on ventilation systems. The AP600 control room habitability design criteria should not be any less robust than the i criteria for current operating reactors, or advance designs such as the l ABWR and CE System 80+. At a minimum, the VES should be designed in :
accordance with the American Society of Heating, Refrigeration and Air- ;
conditioning. Engineers, Inc. (ASHRAE) which is a widely accepted '
authority in the field of HVAC engineering.
1 Therefore, the staff finds it acceptable for each train of VES to have the capacity to independently supply air for 11 occupants in the MCR ,
during DBA conditions in accordance with (1) ASHRAE Standard, IC 62-1989 " Ventilation for Acceptable Indoor Air Quality," and (2) the 1993 ASHRAE Handbook, " Fundamentals, SI Edition", Chapter 23.2, " Venti-lation and Indoor Air Quality." SSAR Section 6.4 should be updated accordingly.
SSAR Section 6.4.5.3 states that connections are provided for sampling the air supplied from the compressed and instrument air system and for '
l periodic sampling of the air stored in the storage tanks. However, Westinghouse should state that the samples are tested in accordance with the accepted air quality standards and specify the current indus-try standards for bottled air quality acceptability and revise SSAR Section 6.4.5.3 accordingly.
l l (8) Sealing Materials . j RGs 1.52'and 1.140 states that silicone sealants or any other temporary l patching materials on filters, housing, mounting frames, or ducts should not be allowed. The reason for not permitting sealant materials ,
is that' unfiltered in-leakages may enter the MCR envelope through duct l and filtration equipment outside MCR envelope and through common ;
. interfacing barriers to the MCR envelope boundary if sealants are cracked or degr ded.
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In spite of this guidance, the AP600 design incorporate common sealants without adequately establishing the environmental qualification for the sealant materials. Even though technical specification tests are performed which are intended to demonstrate acceptable performance of the VBS, the use of these sealants may yield erroneous test results.
Also, certain design weaknesses may give additional erroneous test results. For example, faulty or cracked sealants used on non-safety-related HVAC system ductwork routed through the MCR envelope may lead to unfiltered in-leakage that could provide false indications of pressurization of the MCR envelope.
l Additionally, Westinghouse states that the steam from the SFP will not adversely impact the environment for-the sealing materials because the steam is vented from the fuel handling area directly to the atmosphere j from a vent located more than 150 feet from the MCR. Westinghouse should provide, for staff review, a detailed evaluation of the environ- .
mental effects on the MCR sealant material from the onset of spent fuel !
pool boiling until the blowout panel is lifted. The appropriate SSAR !
Sections should be updated accordingly.
1 (9) Instead of relying on natural circulation cooling which would require l complex analyses to demonstrate acceptabf 9ty, Westinghouse now '
propose: to provide MCR and I&C room cociing during post-72 hour scenarios with ancillary fans which should maintain temperatures close i to the average outside air temperature. The staff finds this proposed I change satisfactory provided that acceptable performance of the ancil- !
lary fans can be demonstrated via inspection, test, analysis, and acceptance criteria (ITAAC) and the initial test program. 1 1
(10) Technical Support Center (TSC) Habitability Requirements l Westinghouse states that the habitability of the TSC under accident l conditions with offsite or onsite ac power available will be equivalent '
to the MCR as described in SSAR Section 9.4.1. SSAR Sec-tion 9.4.1.2.3.1 states that "The main control room / technical support center HVAC equipment and ductwork that form an extension of the main control room / technical support center pressure boundary, limits the overall infiltration (negative operating pressure) and exfiltration (positive operating pressure) rates to those values shown in Table 9.4.1-1. Based on these values, the system is designed to maintain operator doses within allowable GDC 19 limits." Westinphouse should provide dose calculations for NRC staff review to demonstrate the above conclusions and to close OITS Item No. 3085.
For post-72 hour worst case scenarios, the TSC equipment will not be powered and ventilation is not available and MCR communication links with offsite locations would be used to support the MCR staff.
Westinghouse has provided a revised response to RAI 100.10 (OITS Items 1222 and 2033) concerning the unavailability of the TSC. The acceptability of the RAI response is still under review by the staff.
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-(11)- The Westinghouse response to Comment (11) relates to the dynamic effects of the modification of the seismic shield building roof struc- ,
ture model in reflecting the design changes of PCS tank structures and I the increased amount of water to the seismic responses (locally and globally) of the nuclear island structures. Westinghouse states that: ,
- a. The seismic model (3D lumped-mass stick model of the shield build- !
ing roof is being updated to reflect the change in structure and ;
- water inventory. The snow mass will be also included in this ;
revised model. The modeling technique follows the same procedure i used to develop the existing model. When the revised model is ;
complete, comparisons of (a) frequencies and mode shapes, and- .
(b) response spectrum analysis for seismic input at the cylinder base will be made between the existing stick model including water i sloshing mass, and the updated stick model including water sloshing '
mass to determine if the design changes are significant,
- b. In comparing the evaluation results and determining the i significance of the design changes, Westinghouse proposed the :
following criteria:
frequency of significant modes differ by less than Z percent
- translational acceleration response of structural masses in updated model do not exceed those from the existing model by ]
more than 5 percent member forces at the base of the updated model do not exceed those from the: existing model by more than 5 percent
- c. If the global effects due to the design change in the PCS tank and-shield building roof structures are found to be significant from the evaluation in (b) above, Westinghouse will increase the masses of the auclear island model to account for the mass due to live load in accordance with the criteria for live load mass considera- l tion. Comparisons will be made at a few representative locations against the floor response spectra from the existing analyses at 5 percent damping. The change will be considered insignificant, if J
- frequency of significant peaks on the floor response spectra ,
differ by less than 5 percent ]
magnitude of peaks'of updated response spectra does not exceed existing response spectra by more than 10 percent
- member forces from the uodated model do not exceed the existing member forces by more th'n a 10 percent The staff finds Westinghouse's approach acceptable for the modification i of the existing seismic model of the shield building roef structure by ;
(1) revising the member stiffness to account for thinning the inner wall i of the PCS tank, and (ii) changing mass properties to the existing mass i points to account for increase of water inventory and snow load. j i
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- However, Westinghouse intends to address the staff's concern by
- . evaluating the local impact to the roof structure design due to the i design change as described in its responses (a) and (b) above. To consider only the local effects by performing analysis for the roof structural model is not acceptable because, as stated in the staff's letter dated May 1,1997, the evaluation of impact due to the design change of the roof structure should take live loads into account. In j' addition, the use of the existing floor response spectra as input at the i
base for the analysis of the shield building cylinder will not be proper because.these response spectra are to be changed when live loads and
' design changes are taken into consideration. Therefore, the application of the evaluation criteria for the determination of significance of the design change become meaningless.
e As for the Westinghouse's response described in (c) above, the approac'h for revising the existing model of the nuclear island by increasing masses due to design change and live loads appears reasonable to the staff. However, for the case where the global effects due to the design change in the PC3 tank and shield building roof structures are found to be significant from the evaluation in (b) above, Westinghouse proposes to make a comparison based on the results from a hard rock fixed-base 1
seismic analysis against the existing analysis results. This is not acceptable to the staff. The basis for the staff's conclusion is that
> the effect of increasing mass properties to the structural responses may 1 not be significant for the hard rock site condition, but it will be
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pronounced for the soil site condition, especially for soft soil site. ;
As far as the criteria for the determining the significance of the )
design changes described in Items (b) and (c) above, the staff's 1 conclusion is that these criteria appear reasonable provided that the seismic response calculations are based on the entire nuclear island 1 seismic model win proper consideration of live load masses. 1 (12) Westinghouse stated in its response to Comment (12) that it is expected that the process described in the response to Coment (11) will demonstrate that the post-72 hour design changes will have no significant gicbal impact and that it is unnecessary to reanalyze each of the design site conditions. Westinghouse also stated that the impact of the design change on the shield building roof will be included in an update of the local analyses and design calculations for the roof.
Westinghouss's response is not acceptable, because as stated in Com-ment (c), the design changes of the PCS tank will raise the height of the roof structure centroid and, in turn, will affect, especially for a soil site, the rocking and translational responses (seismic shear forces, bending moments and in-structure response spectra) during a seismic event. This kind of effects cannot be evaluated without
, performing analysis with the entire Soil-structure interaction system model. In addition, the live load masses should also be considered when the final seismic model is developed.
(13) The staff. finds the' response to Comment (13) satisfactory.
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- i i 8 i (14) The staff finds the response to Comment (14) satisfactory.
(15) The Westinghouse response to Comment (15) provides detailed availability and testing control recommendations for the COL applicant. Westinghouse !
notes that the post-72 hour equipment will be captured ur. der the regulatory treatment of non-safety systems (RTNSS) screening process and !
states that the COL recommendations will be added to the RTNSS implemen- l tation report, WCAP-13856. The staff finds the level of detail for the testing and availability controls of the post-72 hour systems satisfac-tory. However, placement of these controls as COL " recommendations" in !
the RTNSS WCAP does not achieve the level of confidence the staff needs to ensure that the COL applicant will enact the recommended programs. l Westinghouse should place COL applicant commitments for the post-72 hour i availability and testing controls (with the level of detail provided in ;
the response to Comment 15) into the AP600 SSAR. Westinghouse should ,
determine the appropriate sections of the SSAR to place these commit- :
ments. I The staff also needs additional justification concerning the 10 year i frequency of the integrated system testing for each post-72 component 1 (such as comparable industry standard or other common practice )
examples). The staff believes that a refueling outage integrated testing frequency would provide more reasonable assurance of system i overall functionality. J l
In addition, Westinghouse should provide appropriate ITAAC and ITP testing commitments for the post-72 hour equipment.
(16) Westinghouse states that the environmental equipment qualification requirements based on the post-72 hour design changes will be addressed in revision 12 of the SSAR. The staff cannot reach a conclusions about the acceptability of this response until the revision 12 SSAR EQ changes are submitted.
(17) The environmental equipment qualification requirements based on the post-72 hour design changes are addressed in revision 12 of the SSAR.
The staff is currently reviewing this information.
(18) Westinghouse states that all safety related equipment subject to the environmental effects of spent fuel pool boiling are already adequately addressed in the SSAR EQ table for equipment in the fuel handling area which may be exposed to a steam environment. Westinghouse states that no other equipment will be exposed to steam from spent fuel pool boiling since it is vented directly to the atmosphere. The acceptability of >
this response will be evaluated upon Westinghouse addressing the staff's concerns provided in item'(b), " Potential Harmful Environment Generated by Boiling'of the Spent Fuel Pool," above.
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