ML19221A023
| ML19221A023 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/13/1976 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7905160500 | |
| Download: ML19221A023 (9) | |
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50.320 R. C. DeYoung, Assistant directcr for Light later.4eactors, CPt REVISIO:iS TO THE DRAFT SAFETV EVALUATICN REPORT FR Ti! PEE : TILE ISLN;D liUCLEAR STATION, UNIT 2, AND REQUEST FOR ADDITIJ:iAL I:!FOR'4TI1*:
Plant Game: Three Mile Island Nuclear Station, Unit 2 Jocket !!o.:
50-320 Ailestone Jo.: 24-04 Licensing Stage:
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- .SSS Supplier: Babcock & Wilcox srcnit0ct Engineer: Surns & Roe Containment Type: Dry 3esponsicle Erancn and Project. Manager:
L'.6-2; !i. Silver Recuested Cocpletion Date: April 20,1976 Review Status: Awaiting Infomation Enclosed are revisions to the draft Safety Evaluation Report for the Three dile Island Nuclear Station, Unit 2 (TMI 2). This report nas been prepared by the Containment SysteI;s Branch after naving reviewed the aorlicable acrtions of the FSAR including ;c.endments 1 tnrough 39. Additional infor.ation is necessary befort: we can conclude on the adequacy of the containrent functional design, the acceptability of the proposed use of ti;e contain'ent purce during nomal operation, and the acceptability of the minirva containment pressure analysis. Tiie outstanding items are triefly described below:
1.
Containnent Subcencart:aents Analysis The applicant has not justified that pix breaks in the shield wall pipe penetrations need not be analyzed. Also, the applicant nas not provided tae resultant loading on tne reactor cavity structures and conrarad it to design capability.
2.
Liain Steam Line Break Accident Analysis The applicant has not identified all sources of mass and energy which could contribute to the release to the containment.
In addition, the applicant has not adequately discussei the secondary systen isolatien signal (s) and total elaosed tice including instrumentation delay tire for autenatically terminating :nass and energy addition to the affected steat:1 generator.
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3.
Containment Furning Durin7 :;or al Plant l'reration The applicant has indicated tnat containcent purging durina norm 1 operation will be necessary.
It is cur position tnat centainment purge systa::s wnich do not meet our design Suidelines for an on-line purge system, as stated in Branch Technical Fosition CSS 6-4, should be liraited to acout 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year of operation during normal plant oceration. The applicant has cemitted to a Technical Spect-fication liciting purging to less than 90 nours per year (atout 1 percent of the ti=c) during nor-ul plant oceration. This ratter is now resolved.
4.
Heat 2eroyal Systens The MSSS vendor has reanalyzed the containment spray systm cerfor-anc?.
The analysis indicates that the sodium hydroxide tank (SHT), sodiur: thic-sulfate tank (STT), and borated watsr stcrage tank (BWST) wuld not drw down together as previously predicted. This muld result in the eroty-ing of tne SHT and STT up to twenty-two minutes before the Ai:ST is developed. ile will require the applicant to evaluate the effect of uneven drawdown on systam performance including the potential for pump cavitation.
5.
flinfrum Containrent Pressore Analysis The applicant references Topical Report S'J-10103, "ECCS Evaluaticn cf B&W's 177-FA Lowered loop MSSS" for its ECCS evaluation. We have previ-ously requested the applicant to orovide a co-parison of the actual containment paraineters for THI 2 with those used in BA'.'-10103 Lut the applicant has not provided this information.
In addition, the asplicant's analysis implies that only one spray train is used instead of tw scra/
trains. We will require the applicant to cor pare the TMI containnent parameters for the ECCS evaluation to the corresponding parameters in SAW-10103.
All of the above matters have been discussed with the applicant. Enclosed is a request for additional information and a draft Safety Evaluation for these matters that have been resolved. We will completa our review for the recaining outstanding items after receipt of this information.
Robert L. Tedesco, Assistant D'irector for Plant Systees Division of Systems Safety 120 \\ %
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RE0 VEST FOR ADDITIOrlAL INFORMATION (CONTAIhMENT SYSTE4S)
THREE MILE ISLM D NUCL:.AR STATION, UNIT 2 COCKET NO.
50-320 042.0 CCNTAINMENT SYSTEMS BRANCH 042.17 Provide the following information for the reactor cavity analysis:
(6.2.1) 1.
Page 53-42-3d of Supplement 3 to the FSAR indicates that the neutron shield is designed to withstand the differential pressure that may develop across it in the event of a pipe break accident. Provide an analysis of the differential pressure across the neutron shield, and ccmpare the results to the design capability of the shield.
2.
Provide the resultant loadings on the reactor cavity structures and compare them to design values.
3.
Provide justification for not postulating pipe breaks in the reactor shield wall pipe penetrations.
'Je note that information pertaining to the shield plugs, which will no longer be used in the TMI 2 design, has not been removed from the FSAR.
Since we do not agree with the analytical model presented in the FSAR regarding the renoval of the shield plugs under postulated accident conditions,and since they will not be used, any information related to the shield plugs :hould be deleted frca the FSAR.
042.18 The response to 042.7, regarding the main steam line break accident, (6.2.1) discusses three postulated accident cases including:
(a) steam line break with #ailure of the turbine stop valves to close; (b) steam line break with failure of a main feedwater centrol valve to close; and,
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(c) steam line break with failure of an emergency feedwater valve to close.
The following additional infomation is needed in order to ccmplete our review of the main steam line break accident:
1.
For each of the above casas, discuss how reverse steam ficw and/or feedwater additicn to the affected steam generator is teminated; 2.
Discuss and justify the methodology and assumptions used to calculate mass and energy addition (steam and/or feedwater) to the affected steam generator prior to steam and/or feedwater line isolation; 3.
If cperator action is required to isolate a steam or feedwater line to mitigate the consequences of the postulated accident, justify the assumed time for the operator to take action and isolate the line;
- and, 4.
If valve closure cccurs autcmatically to isolate a steam or feedwater line, specify and justify the isolation signals and the total elapsed time, inc'uding instrument del'./ times, for valve closure to cccur.
Discuss the design criteria for the safety system.
5.
Explain what is meant by latching the emergency feedwater isolaticn valve. Provide justification for assuming that emergency feedwater will not enter the affected steam generator.
6.
It is stated on Page $3-42-7C that the S&W version of the CCNTEMPT code was used for the main steam line break accident analysis. This 120 M0
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@ versicn of the ccde does not include liquid dropcut of the condensate ferred on passive heat sinks when the centainment it csphere is sucer-heated.
It is our position that All cendensate on the Mat sink.; shculd be acded directly to the sump. Provide justification for any other model that is used.
In addition, provida the equatiens and assumptiens used te calculate the mass of ccndensate roulting frca condensat1m on the passive heat sinks for both saturated and sucerheated cend;tio:is in the containment.
7.
Tio response to 042.7(2) has been presented; provide the requested infer-maticn:
042.19 In response to Question 042.10, it is stated that the centainment at. oschere (6.2.4) is monitored for the hydrogen centent by means of a Iccal saecoling staticr.
flo justification is given to shcw that the operator will have encu;E time to analyze the sample and take action before the ccncentration of hydrogen approaches the 4 percent limit. Justify that the proposed sampling technique is an acceptable way to monitor the hydrogen concentration within the centain-ment following an accident.
042.20 The response to 042.16 is unacceptable. Discuss your plans for providing a (6.2.4) purge system that ccmplies with Branch Technical Position CSB 6 4,
" Con ta in-ment Purging During flormal Plant Operations," and provide the analyses identified in Item 3.5 of the Branch Technical Position to justify the contairnent purge system desi5n, o* cemit to limiting carge system operation to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year.
\\23 \\N 042.21 The infomation presented in your letter dated February 26, 1976, rega rd-ing the minimum containment pressure analysis for the ECCS evaluation is inccmplete. Therefore, provide the folicwing infomation:
(a) Provide a comparison between the Three Mile Island, Unit 2, centain-ment parameters and those presented in the Ba'.1 Toofcal Report MW-10103.
(c) Specify the number of spray pumps assumed cperable for the ECCS analysis, the spray flow rate and starting time,and ccc are it to BAW-10103.
(c) Specify the number of. air coolers assumed cperable for the ECCS analysis, the air ecolers duty as a functicn of time and the starting time,and compare it to MW-10103; and, (d) Table III-2 of the acove referenced letter shows that reactor 2
building dome surface area to be 18,650 ft, and Table 6.2-1 in the 2
FSAR shcws this area to be 23,000 ft. Provide the correct surface area.
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RUISI0 tis TO THE DRAFT SAFETY EVALUATI0ft REPORT (C0:1TAItiME:li SYSTE'is)
THREE MILE ISLAfiD STATIO:1, Ui1IT 2 DCCKET ll0.:
50-239 1.
Delete the last two paragraphs of Section 6.2.3 and replace them with the following:
"With regard to the containment purge system, the applicant proposes to intemittently purge the c::ntainment during nomal plant operation.
However, since the system supply and vent lines are larger than that recem. ended in Branch Technical Position CSB 6-4, ' Centainment Purging During lJor al Plant Operations,' we will require the applicant to limit purge system operation to less than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year (abcut 1 percent of the time) and will include this limitation on the use of the containment purge system in the plant technical specificaticns. The applicant has agreed to this ccmmitment."
2.
Delete Section 6.2.5 and replace it with the follcwing:
6.2.5
" Containment Leakage Testing Program The Three Mile Islane fluclear Station, Unit 2, containment design includes provisions and features to satisfy the tescing iequirements of Appendix J to 10 CFR Part 30. However, the containment leak test program contains exceptions to the requirements of Appendix J regard-ing the personnel air lock test precedure and frequency, fuel transfer tube and equipment hatch test frequency and the provisions for correc-t1ve action and retest of local leaks to reduce the overall measured containment integrated leakage rate to an acceptable value. We will require the applicant to revise containment leak testing procedures 120 199
. and test frequencies to comply with the requirements of Appendix J during the preparaticn of the plant Technical Specifications.
"Ccmpliance with the requirements cf Apcendix J to 10 CFR Part 50 provides adequate assurance that containment leak-tigh+. integrity can be verified througnout service lifetime,and that the leakage rates will be periodically checked during service on a timely basis to maintain such leakages within the specified limits - The plant Technical Specifications will contain appropriate surveillance requirements for containment leak testing including test frecuencies.
" Maintaining containment leakage rates within such limits provides reasonable assurance that, in the event of any radioactivity releases within the ccntainment, the loss of the containment atmosphere through leak paths will not be in excess of acceptable limits specified for the site. Ccmoliance with the requirements of Acpendix J constitutes an acceptable basis for satisfying the requirements of General Design Criteria 52, 53 and 54 of Appendix A to 10 CFR Part 50."
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