ML19220B922
| ML19220B922 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/23/1976 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904270662 | |
| Download: ML19220B922 (6) | |
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50-320
. DeYeurg, Assistant Director for Light Water.teactors, "'
.:Ui3T FC:i ADDITIC.ML I:iFLPETI3 ('.;-3) FM T! RIE
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JTATIO.i, UIT 2 Flant riana: Three Mile Island.;uciaar Station, Unit 2 Jcclet io.:
50-323 Stilestene..o.:
24-04 Licansing Stage: CL l1$35 Supolier: Eatcock 3..'ilcox
.ircnitect Cncineer:
Surns 5 'loe Ocntainuent Tg e: Dry
.lcs;onsi:,le Cranch a Project Manager:
L..A, 3 ranch ::o.2, :i. Silver Aequested Ccapletion Jate; deview Status: Awaiting Inferration The enclosed request for additional infomation (Q-3) for the Three ' tile Island iluelear Station, Unit 2, has been prepared by the Contain. ment Systems Branen after having reviewed the apcropriate sections of the Final Safety Analysis.?eport (FS.u), as amended, up to and includin-
/-nendment 33.
The following is the status of issues that reain to be resolved:
1.
The applicant has changed the reacter cavity desi;n to eliminate movable obstructions to vent flor, the anglicant procoses to install a fixed neutron snield structure over the reacter cavity. !!e will require additional analysis and infor ation resarding the :.odeline of tne neutron snield structure in order to cocolete our review and confir-'atory reactor cavity aralysis. The informtion needed is included in tne attached request for additional infor-ation.
2.
The applicant has not responded to a previous recuest for additional
.infor ation (question 21.50) regarding tne nain steam line treak accident.
3.
The acclicant has not ccm.itted to wrfom the contairsent inte; rated leakaqe rate test at the maxinum calculated containment pressura of 55.7 psig as required by Aapendix J to 10 CFR Part 50 We vill requira the applicant to cm: ply with Appendix J.
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.:ita tuse a:suaec by 30.. in El.','-10103.
3.
Ti.e apolicant das advised us t:ut a rajor revision to Table 5.2-15 "Ccatair ant Isolation Valvcc," uill ;e sutritted in a later mena-
.ent.
TNeefore, car cor.clasions in tne Ca fety Evaluatico '.ecort reariinn t1e contain.. ant isolation syste are subiect tc chany.
6.
T% anolicant has not provided an accootable respence to cur incairy into t':0 proccsad u.2 of tne centain ent curge system durin onrral 00 era tion.
- ur pc:ition his been stated to the a:clicant in ivestien 04.16.
. c. ever, the a ?licant has not a.'e mtalv di;cusseI Tc s 9 presant syste desi~n satisfie, cr ches net s tisfy Cracch T civic 21 Position C33 S-4.
The restense appears to cugg::t that unli.ite?
Ourge sj: ten operation will be neces: art'.
!? this is the case, t:.a applicant aas c.ct discussed the plart desien :eculiarities t.,at will necessitate unlir.ited purge syste, operation.
Orir,ini mned D nobert L. Tede5C
?.obert L. Tadesco, e ssistant irtetor for ?lant Systems Division of..jstrs Saf at;-
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'i. Lainas H. Silver S. Varga J. Snacaker J. ".udrick F. 21twila
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REOUEST FOR ADDITIONAL INFOR:L' TION (CONTAI'.7ENT SYSTEMS)
THREE MILE ISLAND '.UCLEAR STATION UNIT 2 LOCKET NO.-
50-320 042.0 CONTAIN!ENT SYSTEMS BRANCH 042.17 Provide the following information for the reactor cavity subcompartnent (6.2.1) analysis:
1.
Provide sufficiently detailed plan and section drawings for several views showing the general arrange =ent o f the reactor cavity struc-tures, components, piping, and other major obstructions such as the proposed neutron shield. These drawings should identify all subcom-partment nodes and flow paths.
2.
Provide and justify the values of the vent loss coefficients and/or friction factors used to calculate flow from the top of the reactor cavity and around the proposed neutron shield.
3.
Page S3-42-3d of Supplement 3 to the FSAR indicates that the neutron shield is designed to withstand the differential pressure that cay develop across it.
Provide an analysis of the differential pressure across the neutron shield, and compare the results to the design capability.
4.
Previde tha resultant loadings on the reactor cavity structures, reactor vessel and vessel supports, and compare them to design values.
We note that information pertaining to the shield plugs, which no longer will be used in the TMI 2 design, has not been re=oved from the FSAR. Since "we do not agree with the analytical model presented in the FSAR to analyze their renoval under postulated accident conditions and since they will not tM - 04 N s
be used, any information related to the shield plugs should be deleted from the FSAR.
042.18 The response to 042.7 regarding the =ain steam line break accident is (6.2.1) ince:plete. Provide the following information:
1.
Identify the equipment and cceponents relied on to limit the mass and energy released to the containment following a main steam line break. Specify the design criteria for this equipment and compon-ents.
2.
For each case analyzed above, identify all suurces of mass and energy and the time periods during which each source is added mass and energy to the containment.
3.
Provide a tabulation of the results of the above analyses, including the =aximum containment pressure and temperature, and tinc(s) of occurrance, and specify the active heat removal equipment assumed to be operable.
4.
Graphically show the containment pressure and temperature responses for the cases analyzed.
042.19 No response to Question 21.50, which also is concerned with the sain steam line break accident, has been provided. Provide a responre to this question.
042.20 In response to Question 042.10 it is stated that the containment atmosphere (6.2.4) is monitored for the hydrogen content by means of a local sampling station.
No justification is given to show that the operator-will have enough tire to analyze the sample and take action before the concentration of hydrogen 84-049
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, approaches the 4% limit. Justify that the proposed sampliag technique is an acceptable way to onitor the hydrogen concentration within the contain-cent following an accident.
042.21 The response to 042.11 is unacceptable since it is not clear what is meant (6.2.4) by the state =ent that another recombiner unit will be stored locally.
Specify whether this hydrogen recombiner unit will be stored on site.
If you intend to share reco=biners between nuclear power plant sites, discuss your plans for transporting the shared reco=biner.
042.22 Describe the instrumentation that will be provided to monitor the perf or nace (6.2.4) of the hydrogen reccroiner to assure that it is performing its intended function. It is our position that such instrumentation sheuld be provided and that readout and alarm capability should be provided in the control room.
042.23 The response to Question 042.15 is unacceptable. Provide justification by (6.2.4) reference to statements in Appendix J that certain containment isola tion valvce (listed in Table 6.2.15) need not be Type C tested.
042.24 Provide the analyses identified in Ites 3.5 of 3 ranch Technical Position (6.2.4)
CSB 6-4, "Contain=ent Purging During Nor=al Plant Operations" to justify the containment purge system design.
042.25 The response to 042.16 is unacceptable. Discuss your plans for providing (6.2.4) a purge syste= that cceplys with 3 ranch Technical Position CS3 6-4 or co=mit to limiting purge system operation to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year.84-050 O
.. 042.26 Section 6.2.1.4.2 indicates that the containment integrated leakage rate (6.2.6) test will be perforced at a pressure of 51.4 psig. Ecwever, in Supplement 2 of the FSAR (Page S2-138) the maximum calculated pressure is reported to be 55.7 psig.
It is our positien that the ILRT should be conducted in cocpliance with Appendix J at the calculated pressure of 55.7 psig.
Discuss your plans for ccmplying with Appendix J.
042.27 Discuss the capability of the containment spray pumps to function reliably (6.2.1) following the emptying of the chemical additive tanks. Describe any tests that have been conducted to verify this capability.
042.28 The information presented in the response to 042.1 regarding the minimun (6.2.1) contain=ent pressure analysis for the ECCS evaluation is unacceptable.
Provide a eccparison between the Three M.11e Island, Unit 2 containment parameters and those presented in the B&W topical report. Justify the applicability of 3AW-10103.
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