ML20072H013
| ML20072H013 | |
| Person / Time | |
|---|---|
| Site: | Satsop |
| Issue date: | 03/24/1983 |
| From: | Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| GO3-83-248, NUDOCS 8303290240 | |
| Download: ML20072H013 (10) | |
Text
.. r Washington Public Power Supply System Br.,x 1223 Elma, Washington 98541 (206)482-4428 Docket No. 50-508 G03-83-248 March 24, 1983 Director of Nuclear Reactor Regulation ATTN: Mr. G. W. Knighton, Chief Licensing Branch No. 3 Division of Licensing US Nuclear Regulatory Commission Washington, D. C.
20555
Subject:
NUCLEAR PROJECT 3 RESPONSES TO NRC ACCEPTANCE REVIEW QUESTIONS (February 1983)
References:
a) Letter D. G. Eisenhut to R. L. Ferguson, dated 08/20/82 b) Letter #G03-82-830 G. D. Bouchey to H. R. Denton, date 08/20/82 c) Letter #G03-82-1085 G. D. Bouchey to J. D. Kerrigan, dated 11/22/82 Reference a) transmitted a set of questions generated during the NRC's acceptance review of the WNP-3 Operating License Application (reference b).
Reference c) represents the initial Supply System response to these questions and provided a schedule for those cases where our evaluations were not yet complete.
i This letter transmits those responses scheduled to be provided for NRC re-view in February. In those cases where it is considered necessary or desir-able to amend the FSAR due to our responses, we have provided marked up FSAR pages which show the changes which will be included in a subsequent amendment.
8303290240 830324 PDR ADOCK 05000508 A
- Mr. G.
W.' Knighton Page 2 March.24, 1983 RESPONSES TO NRC-ACCEPTANCE REVIEW QUESTIONS If you require additional information or clarification. the-Supply System point of contact for _this matter is Mr. K. W. Cook,. Licensing Project Man-ager (206/482-4428 ext. 5436).
Sincerely,
&%w G. D. Bouchey, Manager Nuclear Safety and Regulatory Programs AJM/ss Attachments:
1.
NRC Question 410.8 (9.2.5) 2.
NRC Question 460.1 (11.5.2.4.2) 3.
NRC Question 471.1 (12.3.4) 4.
Request for Additional Information - Enclosure 4 (item 3)
Safety Related Structures. Systems and components (Q-List) controlled by the QA Program.
cc:
D. J. Chin - Ebasco NY0 N. S. 'Reynolds - D&L E. f. Beckett - NPI J. A. Adams - NESCO
2 ATTACHMENT 1 Question No.
410.8 (9.2.5) f Section 9.2.5 does not define the number of cells per cooling tower train, nor does it contain figures 9.2.5-2a through 9.2.5-2d.
Confirm the date by which you intend to supply this information.
Response
As committed to in Letter G03-82-1085 dated Octob
'987, the att>
1 Figures 9.2.5-2a through 9.2.5-2c are provided f heat release curve and sensible heat removal cu-this section (replaced by Shutdown Heat Exchange this data is covered in Subsection 6.2.1 and 5."
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1617W-2 WNP-3 FSAR During normal ~ operation the UHS in conjunction with the CCWS heat exchanger, i
can reject the maximum normal _ heat loads while maintaining CCWS temperature at or below 95F. Table 9.2.5-1 presents expected tower performance requirements during normal conditions. Parametric performance expectations for a range of ambient temperatures with and without heat rejection to the SWS under normal-and emergency conditions is presented in Table 9.2.5-2.
During emergency operation the UHS provides sufficient cooling to setety-related heat loads identified in Subsection 9.2.2.
In the event of the loss of one train for any reason, the redundant train can rey et the maximum instantaneous heat load.
Non essential heat loads can be manually aligned _as reactor decay heat affords spare UHS capacity. The appropriate controls as discussed in Section 7.3 permit the operator to monitor and select these heat load alignments.
9.2.5.3 Rfety Evaluation
'The UHS, with one tower operating alone is designed to meet maximum heat load following an accident. Its long term heat rejection capacity is sufficient to mitigate a postulated LOCA and return the containment pressure and temperature to ambient conditions.
The results of an analysis of the 30-day period following a design basis accident are found in Tables 9.2.5-4 and 9.2.5-5 and Figures 9.2.5-2a through k 9_.2.5-2 6 This analysis has determined the total heat rejected, the sensible l2 heat rejected, the station auxiliary system heat rejected, and the decay heat release from the reactor)(.beklls f &:S w hs4 art (ran4J. N b ha*5 6.2. f f 5.4.7.
'g As per Branch Technical Position ASB-9-2, the decay heat curves for fission products and for heavy elements were obtained using the assumptions and uncertainties set forth in the October 1973' draft proposed ANS standard,
" Decay Energy Release Rates Following A Shutdown Of Uranium-Fueled Thermal Reactors" (ANS-5), to establish the heat input due to decay of radioactive material. An equilibrium fuel cycle and an increase in the calculated heat inputs were assumed as follows:
3 a) hor the time interval 0 to 10 seconds, 20 percent was added to the heat released by the fission products to account for uncertainties in their nuclear properties.
3 7
b)
For the time interval 10 to 10 seconds, 10 percent was added to the heat released by the fission products to account for uncertainties l
in their nuclear properties.
7 c)
For the time interval 0 to 10 seconds, the heat released by the heavy elements was calculated (using the best estimate of the production rate for each unit) and 10 percent was added to account for uncertainties in their nuclear properties.
I 2
9.2-45 Amendment No. 2, (12/82)
N' 400 4000 200 2000 INTEGRATED HEAT REMOVED BY HEAT EXCHANGER 3
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HEAT REMOVAL RATE 6
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MSPRAYS = 2.7 x 10 LBM/HR 0-0 0
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UNFOULED SHUTDOWN HEAT EXCHANGER ER LY SYSTEM HEAT REMOVAL2 (U = 302 BTU /HR - FT oF) 9.2.5-2a Nuclear Proj.ects 3 & 5 DESLS WITH MAX St FINAL SAFETY ANALYSIS REPORT
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1 AUX. HEAT REJECTION RATE - 10 BTU /HR 0
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105 10 107 6
103 10 TIME (AFTER LOCA) - SEC FIGURE WASHINGTON PUBLIC POWER SUPPLY SYSTEM HEAT REEC ION DUE TO STATION AUXILIAR_Y SYSTEMS *
' 9.2.5-2b i
Nuclear Proj.ects 3 & 5 FINAL SAFETY ANALYSIS REPORT
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3 10 TOTAL INTEGRATED HEAT REJECTED 8
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TIME (SEC) l WASHINGTON' PUBLIC FIGURE l
POWER SUPPLY SYSTEM TOTAL HEAT LOAD FOR COMPONENT COOLING. VATER SYSTEM 9.2.5-2c Nuclear Proj.ects 3 & 5 FINAL SAFETY ANALYSIS REPORT l
ATTACHMENT 2 Question No.
460.1 (11.5.2.4.2)
Supply information relating to the effluent radiation monitors for steam generator blowdown flash tank vent and steam seal gland steam condenser
- ventilation which the FSAR indicates as later or provide a schedule for sub-mittal of this information.
Response
In letter G03-82-1085 dated October 22, 1982, the Supply System committed to provide the sensitivity for the Steam Generator Blowdown Flash Tank Vent Radiation Monitor, Auxiliary Condensate Flash Tank Radiation Monitor, and the Steam Gland Seal Steam Exhaust Radiation Monitor.
This information is still unavailable. We expect to provide it by April 1983.
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Question No.
471.1'
- (12.-3.4 )
As per Regulatory Guide -1.70 indicate whether, and if so how, the guidance provided by = Regulatory' Guide -1.97 has been followed concerning area radia-tion and airborne radioactivity monitoring ' instrumentation.
Reference or-provide this information.
Response
.The WNP-3 design for post-accident area radiation and-airborne radioactivity monitoring instrun.entation complies with USNRC Reg. Guide 1.97 P,ev. 2 re-quirements with the. following exceptions; (1), Instruments monitoring _ radiation exposure rates inside buildings or -
4 i
areas which are.in direct contact with primary containment, where penetra-tions and hatches are located (Type C variable) will' not be' supplied.
An increase in radiation levels in-these areas would be due primarily to~
~ streaming through the penetrations or to direct. shine from the containment caused by elevated exposure rates inside the containment.
Under these con-ditions, any additional increase-in the radiation levels as a result of airborne radioactivity leaking from the containment would not be discrimi-nated from streaming or direct shine by the' area radiation monitors recom-rended by RG 1.97 Rev. 2.
The four Auxiliary Building Airborne Radiation Monitors provided in the Reactor Auxiliary Building can detect airborne radioactive material leaking from the containment.
(2)
In order to comply with the requirements for Area Radiation monitoring (Type E variables) in areas where access may be required to service or operate equipment important to safety, monitors have been placed at loca-tions of WNP-3 which were identified and analyzed as vital areas _in the WNP-3 THI Shielding Study (See FSAR APP.-12A). All the dynamic range recommended in RG 1.97 Rev. 2 (10-pnitors supglied have R/hr to 10 R/hr) with the exception of the monitors supplied in the two Valve Operating En-closure areas 'at elevation 351.00 ft of the RAB. According to the TMI Shielding Study the maximum dose rate expected in these areas aft'er an' 2
accident is not m9re than 5 R/gr; therefore, the supplied monitors have a dynamic range 10- mR/hr to 10 mR/hr.
The Supply System is considering the addition of an area. radiation monitor for the Post Accident containment atmosphere monitoring and sampling area.
and the monitor would have a range of 10 pis area is agpxoximately 0.7 ~R/hr The expected post accident dose rate in t mR/hr to 10 mR/hr. The staff will be informed as to the decision to add this monitor by June 1983.
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ATTACHMENT 4 REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4 3)
SAFETY-RELATED STRUCTURES, SYSTEMS AND COMP 0NENTS (0-LIST) CONTROLLED -
BY THE QA PROGRAM Staff requests for additional information regarding this issue have been sent to a number of OL applicants. A request from the Diablo Canyon review is provided as Enclosure 5.
Response
In our original response to this question (letter #G03-82-1085) we stated that the Supply System was in the process of developing the Q-list. We expected to have finalized this list by February 1983.
Due to the amount of time required to process this request we must extend our earlier schedule date for provision of the Q-list to June 1983.
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