NRC-95-4414, Forwards Three Copies of Westinghouse Responses to NRC Open Item on AP600 from 940818 & s,Rev of Response & Listing of NRC RAI

Forwards Three Copies of Westinghouse Responses to NRC Open Item on AP600 from 940818 & s,Rev of Response & Listing of NRC RAI

ML20081D439
Person / Time
Site:	05200003
Issue date:	03/13/1995
From:	Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Borchardt R NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
DCP-NRC0286, DCP-NRC286, NTD-NRC-95-4414, NUDOCS 9503200240
Download: ML20081D439 (12)

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Westinghouse Energy Systems Ba ass r

Pittsburgh Pennsylvania 15230 0355

_ Electric Corporation NTD-NRC-95-4414 DCP/NRC0286 Docket No.: STN-52-003 March 13,1995

' Document Control Desk U.S. Nuclear Regulatory Commission i

Washington, D.C. 20555 i

I ATTENTION:

R.W.BORCilARDT j

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SUBJECT:

WESTINGHOUSE RESPONSES TO NRC OPEN ITEMS ON THE AP600

Dear Mr. Borchardt:

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i Enclosed are three copies of the Westinghouse responses to NRC open items on the AP600 from your i

letters of August 17,1994, August 18,1994 and August 29,1994. In addition, a revision of response i

previously submitted is provided. A listing of the NRC requests for additional information responded to in this letter is contained in Attachment A.

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These responses are also provided as electronic files in Wordperfect 5.1 format with Mr. Kenyon's-copy.

If you have any questions on this material, please contact Mr. Brian A. McIntyre at 412-374-4334.

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W N

Nicholas J. Liparulo, lanager Nuclear Safety Regulatory and Licensing Activities

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B. A. McIntyre - Westinghouse 1

T. Kenyon - NRR

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2(lOOGJ 9503200240 950313 PDR ADOCK 05200003 1

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NTD-NRC-95-4414 i

- ATTACHMENT'A

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AP600 RAI RSPONSES SUBMITTED MARCH 13, 1995-

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440.250 471.22 i

480.17R1

-i 952.93 952.100,

952.102 1

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NRC REQUEST FOR ADDITIONAL INFORMATION 1[EEN nu_

Ouestion 440.250 What is the status of the PRHR inlet line high point trap? If the PRHR inlet line high point trap has been removed, provide new piping schematics.

Response

The passive residual heat removal heat exchanger inlet line high point trap has not been removed. As a result, the piping schematics have not been revised.

SSAR Revision: NONE 3 WB5tingh0USB

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NRC RE' QUEST FOR ADDITIONAL INFORMATION

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=i Question 471.22 This is to formalize a request for additional information discussed during a telecon in June 1994.

a.

The staff's confirmatory shielding calculations confirmed, for the most part, the radiation zone levels described for the AP600 design during normal and accident conditions. However, these shielding calculations indicated rather high (approximately 95 Rem /hr) post-accident radiation levels in the PASS sample room. Determine the time it will take the operators to perform required post-accident actions in all vital areas (as required by 10 CFR 52.79(b) and described in item II.B.2 of NUREG-0737), and provide estimated personnel doses for each of these activities for the totallength of the accident.

b.

Justify why the remote shutdown workstation is not considered a vital area.

Response

a Conunitment to the requirements of 10 CFR52.79(M relative to vital area access and post-accident sampling (10 CFR 50.34 Items (2)(viii)) is included in Section 1.9.3 of the SSAR. He access routes for the various post-accident actions in vital areas have been provided in SSAR Figure 12.3-2. Time estimates have been made for ingress, egress, and performance of actions at the vital area location for each of the post-accident functions.

These times are considered with the post-accident dose rates at the various locations along the access routes and at the vital areas, when determining the total integrated dose to an operator for each post-accident action. The elapsed time from the accident to the time that the action is performed is considered in determining the radiation environment at the locations of interest. The access times and integrated dose to an individual for each post-accident action are listed in Table 471.22-1. As indicated in this table, the maximum integrated dose for each of the actions satisfies the NUREG-0737 requirement of less than 5 Rem whole body, or its equivalent to any part of the body of an individual, for the duration of the accident.

1 The maximum individual dose associated with post-accident sampling are designated as "less than 5 Rem" in Table 471.22-1. Additional explanatory information relative to the projected personnel dose at this location is presented below:

The dose rates and access times for the Post-Accident Sampling System (PASS) sampling activity are included in Table 471.22-2, which reflects a conservative assessment of the radiation environment associated with the

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access and sample procurement activities. For example, dose rates associated with the sampling activities are based on a time after the accident that corresponds to the peak dose rates inside the auxiliary building. Since j

there is a waiting period for sample line flushing, the operation is assumed to be performed by two separate crews; that is, an initial crew to initiate sample line flushing and a separate crew to procure and remove the i

sample. The individual dose, as well as the collective dose to all personnel involved in the operation is expected i

to be much less than the value of 4.93 Rem that is indicated in Table 471.22-2. This is because

1) Since the :ampling function is not safety related and exposure to the public is not an issue in the analysis, more realistic assumptions relative to the activity distribution within the auxiliary building are warranted.

3 Westinghouse m.224

NRC REQUEST FOR ADDITIONAL. INFORMATION ine v

More realistic assumptions reduce the calculated radiation environment to levels such that the projected individual dose associated with the completion of the sampling activities will be much less than 5 Rem.

2) Since no safety-related credit is taken for post-accident radioactivity sampling, some flexibility in the sampling schedule or deferral of sampling is expected to be possible if the magnitude of the post-accident dose rates approach those that are assumed in the analysis.

3) If the measured dose rates are such that the sampling activities, including the waiting period for sample line flushing, can be completed by a single work crew without exceeding the dose criteria, the individual (s) need not be replaced by a separate work crew. This would result in less total personnel exposure associated with the sampling operation.

The calculated dose rates are conservatively based on an instantaneous transfer of all containment leakage into the auxiliary building, with no hoidup or dilution in the annulus regions between the containment wall and the inside of the auxiliary building and no holdup or dilution between floors of the building. A recent EPRI-sponsored study (Reference 471.22-1) concludes that the so-called " middle annulus" provides a significant reduction factor for noble gases that exit this volume through simple holdup (and associated radioactive decay). For example, Time After LOCA (Hours) 8 24 Total Noble Gas injected into middle annulus (grams) 380 1750 Noble Gas Released -single node-100% / day (grams) 50 600 Ratio (Injected / Released) 7.6 2.9 or an effective reduction factor of at least 3 for noble gases. Since noble gases are the major source of exposure to personnel inside the auxiliary building, a reduction factor of 3 to the dose estimates results in a maximum individual dose of 2.2 man-rem for a single crew, which is well below the dose criteria of 5 rem to an individual.

Further, the dose rates are based on complete holdup of the activity within the auxiliary building, with no credit for releases from the building. llence, consideration of more reasonable assumptions relative to the activity transport can be expected to result in even lower dose rates in the occupied areas. Based on the above i

considerations, the personnel dose associated with post-accident sampling is pro;ected to be substantially less than 5 Rem to any individual.

b.

As stated in Section 7.4.3.1.1, the remote shutdown workstation "is designed to allow control of a shutdown following an evacuation of the control room, coincident with the loss of offsite power and a single active failure. No other design basis event is postulated." The design basis for the remote shutdown workstation is i

provided in SSAR Section 7.4.3.1.3 and, as specified, the AP600 design conforms with Reference 471.22-2.

Since no design basis events are postulated beyond the scenario resulting in evacuation of the main control room, no emergency habitability provisions (including post-accident radiation protection and shielding measures) are provided for the workstation area. The design provisions incorporated in AP600 to protect the habitability i

I m22-2 W Westinghouse I

NRC RE' QUEST FOR ADDITIONAL INFORMATION 3

r of the main control room are discussed in SSAR Section 9.4.1.2.4 for the nuclear island non-radioactive ventilation system and Section 6.4 for the main control room emergency habitability system.

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References:

1 471.22-1

" Passive ALWR Secondary Building Mixing and Leak Rate Monitoring", Prepared by Stone &

Webster Corporation for the Advanced Light Water Reactor Program", April 1993.

471.22-2 ANSI 58.61983 " Criteria for Remote Shutdown for Light Water Reactors" SSAR Revision: NONE l

471.22-3

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1 NRC REQUEST FOR ADDITIONAL INFORMATION Table 471.22-1 Personnel Exposure For Post-Accident Actions Post-Accident Action Time After Duration Integrated Dose Accident (hrs)

(min)

Per Individual (Rem)

Main Control Room Access 0

continuous 0.2 - 0.5(l)

Conduct Post-Accident Sampling 28 45 less than 5 Provide Spent Fuel Pool Cooling 64 25 0.77 Make-up Provide Containment Water Inventory 64(2) 52.5 4.69 Make-up Temporary Water flookup to Passive 64 37 0.54 Containment Cooling System Tank Provide Temporary llVAC to Main Control 64 19 1.30 Room Provide Temporary llVAC to PAMS 64 33 1.58 Cabinets Provide Temporary Power to Class 1E 64 48 2.22 Regulating Transformers Provide Main Control Room Air 64 49 1.70 Make-up Note (1) - Dose will depend on mode of operation of the HVAC (see RAI 470.9, revision 1)

Note (2)- The assumption that the action is performed at 64 hours7.407407e-4 days <br />0.0178 hours <br />1.058201e-4 weeks <br />2.4352e-5 months <br /> after the accident is highly conservative since it is estimated that, with design basis leakage, the action would not be required until approximately one month after the accident when the work area dose rates are much lower.

471.22-4 T Westinghouse

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I NRC REQUEST FOR ADDITIONAL INFORMATION ut tts IE IN Table 471.22-2 Access Times and Personnel Exposure Per Individual For Post-Accident Sampling Average Crew No.1 Crew No. 2 Dose Activity Rate Time in Integrated Time in Integrated (R/hr)

Radiation Field Dose Radiation Field Dose (min)

(Rem)

(min)

(Rem)

Proceed to Auxiliary Building entry door 1.67 7.75 0.215 7.75 0.215 Enter Aux. Bldg. door & proceed to stairs 16.3

.25

.068 Descend stairs to Elev. 66'-6' 15.7 2

.522 Proceed to sample room 13.3 0.5

.111 Monitor conditions, align valves & initiate 11.6 10 1.933 sample line flush Exit sample room & proceed to stairs 13.3 0.5

.111 Ascend stairs to Elev.100'-0" 15.7 2

.522 Exit stairs & proceed through rail bay 16.3 0.25

.068 Exit Building & proceed to Secunty 1.67 7.75

.215 Wait for crew #1 outside of Aux. Building

.994 15.5

.257 Enter Aux. Building & proceed to stairs 16.3

.25

.%8 Descend stairs to Elev. 66'-6" 15.7 2

.522 Proceed to sample nmm 13.3 0.5

.111 Monitor conditions, obtain sample

!!.6 13 2.513 Exit sample room & proceed to stairs 13.3 0.5

.111 Ascend stairs to Elev.100*-0*

15.7 3

.783 Exit stairs & proceed through rail bay 16.3 0.5

.135 Exit Budding 1.67 7.75

.215 TOTALS 3.77 4.93 471.22-5 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

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l.b Response Revision 1 Question 480.17 External Film Pattern / Water Distribution Tests Provide additional information on the external film / water distribution that is expected for the AP600. The Waltz Mill tests were done using a steel shell at ambient temperature. Will the film pattern be affected by heating of the shell? Is there a difference in the film behavior in the large scale test facility in cases where the shell is not heated, versus cases where the shell is heated?

Response: (Revision 1)

The SSAR containment analyses assumed the containment wetting increased from 40% at the top to 70% over the outer portions of the dome and the side walls. These wetting fractions were determined from the Water Distribution System Test - Phase 8. Reference 480.17 2, attached lto response"fevisiod 0, presents the results of calculations

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with WGOTHIC which show that pressures within containinent remain within acceptable limits for a case with wetting from 20% on the dome to 40% on the side walls. RefereriEe :480.1733 pro 91de5' additional sensitivities to water coversge to c6nfirm'that.'the AP600 islwell assyy from a rapid change;in the pressure response versus the amount of water coverage The full scale (cold) Phase 3 wetting tests at Waltz mill and the 1/8 scale heated tests are complete? be9 engeing.

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After cen:p!::!ca cf 1 :: :::::, addi: ena! The~information for the hot AP600 coverage and the acceptance limits for coverage [includinicold t6 hotTeffect{was wWweided to the NRC vis ReferenEc}80[17-48 : rev.ced RA! ' ^.ugu :, '99 *

References:

L-:::, F 1. Liparu!c '^ " "'. Berchar& (PC), " ^.96M D::!gn nd D: !gn C ~iF.=tien T::: Program Overien ", Table 3, Rev::!cc 3, Augu~ 13, '993 480.17-2.

M. E. Wills, D. L. Paulsen, V. Notini, G. InvermJi, " Effectiveness of External Cooling and Associated Studies on E AP600 Passive Plant", INC Conference, Toronto, October 1993.

1 480.17-3.;

}Letterj NTDjNRC-94 4286pSupplemental InformationTon ^AP600? PCSJ Filmj FloKCoVerage hfethodologyn Augus_t 31,L.1994

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l Letter NTD-NRC-94424K"AP600 PassiviContainment CoolhtgLSpitenf Lettei Rep 6rts,]ulf 480.174 5 28(1994)

SSAR Revision: NONE PRA Revision: NONE l

i 480.17(R1)-1 W W85tlDgh00S8

NRC REQUEST FOR ADDITIONAL INFORWIATION lir I

L Question 952.93 Provide a commitment to submit a revised CMT Scr. ling Analysis. A CMT test facility scalmg analysis,(WCAP-13963, " Scaling Logic for the Core Makeup Tank Test") was submitted in February 1994, and discussed with Westinghouse on March 14, 1994. Staff comments on the scaling analysis should be incorporated into a revised document which will be submitted to the staff for review.

Response

Reference 952.93-1 was submitted via Westinghouse letter NTD-NRC-95-4390, dated January 31,1995. Revision 1 of the scaling report addresses staff comments on Revision 0 of the report.

Reference:

952.93-1 WCAP-13963, Revision 1. " Scaling Logic for the Core Makeup Tank Test,"

SSAR Revision: NONE 952.93-1 westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

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Ouestion 952.100 Provide a commitment to submit a complete PCCS scaling analysis. The report should describe how the scaling analysis will be used in validation of the analysis codes. The scaling report should describe how the large scale tests analysis will translate to the AP600 design. An explanation of how this analysis would be used in the code validation process should be included.

Response

Reference 952.100-1 provides the scaling report and was submitted via Westinghouse letter NTD-NRC-941318, dated October 27,1994.

References:

952.100-1 WCAP-14190, " Scaling Analysis for AP600 Passive Containment Cooling System,"

SSAR Revision: NONE 1

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NRC R$ QUEST FOR ADDITIONAL INFORMATION at t

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Question 952.102 Provide a commitment to submit an analysis of air flow in the annulus (bo L wet and dry cases).

Response

j An analysis of the air flow in the annulus for both wet and dry condition has been provided in Section 7.2 and 7.3 of Reference 952.102-1. Reference 952.102-1 was submitted via Westinghouse letter NTD-NRC-94-4318, dated l

October 27,1994. Reference 9$2.102-2 provides supporting information for use of a forced convection correlation j

in the annulus.

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References:

952.102-1 WCAP-14190, " Scaling Analysis for AP600 Passive Containment Cooling System,"

952.102-2 Letter NTD-NRC-95-4397, Supporting Information for the Use of Forced Convection in the AP600 PCS Annulus, Dated February 16,1995

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