ML19224C162
| ML19224C162 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/24/1975 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | Schencer A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111230347 | List: |
| References | |
| NUDOCS 7906290227 | |
| Download: ML19224C162 (4) | |
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Denwood F. Ross, Assistant Director for Reactor Safety, DSS B 6 W GENERIC ISSUES During the review of BSAR-241, a number of generic issue.; were identified and transr,itted to B G W with the first round of questions. Since then the BSAR-241 application was withdrawn.
Because a rtain of these issues were applicable to a number of B 5 W plants under review, we initiated a request to B G W (1 and 2) for the subuissicn of topical repolts or other for=s of information suitable for generic This request has been responded to in a series of letters,(review.
N and for the cost part, unsatisfactorily.
With BSAR-205 just docketed and un kr review, it appears that these issues can be best pursued within the review franework of BSAR-205.
The solution of these issues requires the cooperation of cognizant reviewers of both Reactor Safety and Pl:mt Systen.s, and I suggest that review efforts in these areas be initiated in conjunction. with SSAR-205 Review Schedules. The cognizant BSAR-205 reviewer for EIGCSB is R. Fitzpatrick. We request that you appoint a Ley person within your group for appropriate coordination.
OriginM si ned by:
DISTRIBUTION
&b rt L Nmo DOCKET FILES STN 50-561 NRR READING EIC READING' Robert L. Tedesco, A' *istant Director PS READING for Plant Systems R. TEDESCO Division of Syste=s Safety J. GLYNN cc:
T. Ippalito M. Srinivasan C. Miller D. Tondi R. Fitzpatrick D. Basdehas References (Enclosed):
1.
Letter from T. Ippolito to A. Senwencer, dated February 24, 1975.
2.
Letter from A. Schwencer to 2,57 025 J. Malloy dated April 4, 1976.
3.
Letters from K. E. Stdirke to A. Schwencer dated May 23, 1975, July 29, 1975, August 13, 1975, O.
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APR 0 8 576 Denwood F. Ross, Assistant Director for Reactor Safety, DSS B 5 W GENERIC IS3UES During the review of BSAR-241, a number of generic issues were identified and transmitted to B 5 W with the first round of questions.
Since then the BSAR-241 application was withdrawn.
Because certain of these issues were applicable to a number of B 6 W plants under review, we initiated a request to B 5 W (1 and 2) for the submission of topical reporti'or other forms of information suitable for generic This request has been responded to in a series of letters,(teview.
- 1 and for the most part, unsatisfactorily.
With BSAR-205 just docketed and under review, it appears that these
.ssues can be best pursued within the review fra. ework of SSAR-205.
The solutien of these issues requires the cooperation of cognizant reviewers of both Reactor Safety and Plant Systers, and I suggest that review efforts in these areas be initiated ir conjunction with BSAR-205 Review Schedules. The cognizant BSAR-205 reviewer for EISCSB is R. Fitzpatrick. We request that you appoint a key person within your group for appropriate coordination.
EEsw Robert L. Tedesco, Assistant Directir for Plant Systems Division of Systems Safety cc:
T. Ippolito M. Srinivasan C. Miller D. Tondi R. Fitzpatrick D. Basdekas References (Enclosed);
1.
Letter om T. Ippolito to l
A. Schwencer, dated l
February 24, 1975.
2.
Letter from A. Schwencer to J. Malloy dated Aprilg4, 1976.
I 3.
Letters from K. E. Suhrke to A. Schwencer dated May 23, 1975, July 29, 1975, August 13, 1975, and December 30, 1975.
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,4 c-UNITED STATCS NUCLEAR REGULATORY COMM!ssf0N Yv ASHIN G T O N. D.
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20555 i.i FEB t 4 S75 A. Sch:encer, Chief, Light Water Reactors Branch 2-3, frdR REQUEST FOR S&W LICENSING ADDITI0inL II:F0i'JMTICH A number of PSARs with B&U HSSS have been revie:ed recently or are presently under revicw.
These incice the 1: orth Anna 3 &
- 4. Surry 3 & 4, BelleTonte 1 & 2, Green.: cod 2 & 3, UPPSS i & 4 and Pebbie Sprines 1 & 2 plants.
Since a concurrent review of B-SAR-24i was in progress, t:e submittec recuests fcC information on generic itcms on the B-SAR-241 revie.. so that we.can ;pply that informatics in our evaluation of individual applications including those citec ecove.
Bau, hct ever, wi thdrew tha e-SAR-2cl application wititaut responcing to our recuests for additicnci info rma ti on.
A number of these questions was ciso asked in the WPPSS 1 & 4 and Pebble Springs i & 2 applications.
I'cwever, the fecision ' as m dc that it r ould be ir.appropri.
to ask questions of a generic nature of an indivicual appiicant.
Nevertheicss, this infer:
ion snoulc be provided by BSW in a timely manner fer cur &/.aw.
Enclosc; is a list of tocic; for which we need codi.icr.'.! inf erastion.
Thi s i nf or., ti on be submitted in the form o f topical rexrts by the enc of a year cod be ap :licable to all Ea'..' plants curr< ntly in the CF and post-CP stage of review.
He suggest ths t B&'.' sum 1i t ' G Lailed outlines of tha proposed topical reports or other forms of addi;ionai information Mi tnin six. eeks and that a reetina be held bat :cen B&W and the staff to discuss them and arrive al a mutually vceable scope of these topical reports.
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Thomas A. Ipoblito, Chief Electrical, Instrument.nion &
Control Systems Brancn Office of Muclear Reacter Regulation Encloture:
List of Topics cc:
F. Rosa (D. Basdekas
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TOPICS REQUIRII:G ADDITI0flAL IrlF0iUMTIO ;
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FROM B&W 1.
Integrated Control Systep It is stated in all license applications involving a B&W flSSS that the ICS is a non-safety system.
In some applications, the results of an accident analysis perforacd for the ICS are presented.
Submi t additional information regarcing this analysis to show the following for each case analyzed.
(1) The assumptions used and their justification.
(2) The worst case conditions and their bases.
(3) The specific subsystem of ICS involved in each action, the effects of its f.; lure or malfunction on other subsystem of ICS, and tne ultimate effect on safety of the plant.
(4)
Interfaces '..i th systbas' not s' ppli ed by Br..' and the specific u
criteria for these interfaces if related to safety.
(5) A complete list of functions perforced by the ICS in all rnodes of opera tion.
2.
For e.ch of the accidents analyzed in Chapter 15.0, prcvide a detailed listing of all systems and equirnent for which you take credit in making your assumptions in eacn analysis.
3.
For the case of a stem line break Accident inside containment it is a:sumcd that the mai steam line isolation valves and the fced.ater isolation valves to the cffected stcam cenerator close.
Demonstrate that in the event of a steam line break inside containment concurrent with loss of offsite power, no single f.ilure in the main steam and fecdwater isolation valves will defeat these two protcctive actions.
Identifv all instrumentation, cuntrols and electrical equipment that are not designed to Class IE requirements for which credit is taken in this analysis.
Supplement your discussion with a sketch showing the feedwater/ steam system and icentify all coronents that need to be actuated in fil cases of ' feed..'ater/ steam lir.2 break inside end outside containment.
Submi t d.e rasuits of a failure mode and effect analysi.s that demons tra tes that the requirements for feedcrater/ steam line isolation are met while allowing cc,ergency core cooling 4.
Supplement the analysis presented in Sectice 7.4.2 to show that the RPS and CP.DCS designs meet the requirements of GDC 25.
Considerinc recent PUR cperating experience and the fact that the opemtor has provisions to manually centrol single rod motion, the singic rod wi th:'rawal accident should b:' analy.'ed as an anticipa t ed event.
Specifically, provice information justifying the selection of worst-case conditions of time-in-cycle, pc.er level, power dim ribution, peaking factors, control rod worth, control positions, etc.
Justify any differences between these selected values and those used in the rod ejection accident.
257 028 j
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O Provide a plot of maximum fuel centerline temperature as function of time for the control rod misoperation accident.
Demonstrate that the worst case initial conditions have been analyzed.
Justify that the following control rod misoperation events do not have to be analyzed.
(1)
Inaavertent withdrawal of two or more control rods at the same time.
(2) Leaving one or more rods behind (i.e., stuck rods) during rod bank withdrawal.
(3)
Insertion of a rod bank with one or core bank reds stuck.
Show the changes in minir"n DNER and minimum ku/f t for I.he control rod vii thdrawal ana mi soperation eve,r.ts.
, Describe the startup and full pc.:er control cod withdrawal and mis-opera tion ovents.
5.
It is not clear frca the informatien presented in the PSto what yeu-design provisicns are wi th regard to disconnecting nf the Reactor Coolant Pumps from their electric pc.:er bucas becausa uf excessive grid frequercy decay rates, to a nure to.t thc m. n hiretic cuarev is avajirAle f ar ik cm I!c'.m.
Provide ac:iition ' irfor.c.a tiw to shtu ;;hr.t]vour reouire: c..tcc arc in acceicent er.aiyses presented in Chart:r 15.0.
Your respcas shoulo stats the
, limiting f recuency decay rate H /sec.
6.
There is not sufficient information in SAP.s to determine that your design, utilizing tne high and icu pressurizer 1.evel sigr.cl, in lieu of the high contair. ment pressure signc.1, will provic:e an eeuivalent decree of assurance that the reactor will trip prior to or coincident with ECCS actuation.
The staff concludes that if the analysis for the effectiv ness e
of the ECCS perforc.ance takes credit for 6 reactor trip, we require that both diverse sigr.als actucting the ECCS be used to trip the reactor, there, fore, either (a) modify your design to include a high building pressure trip to trip the reactor; or.(b) ce'onstrate that a hign and icw pressurizer level signal will perform satisfactorily for all accident conditions, ar.d will trip the reactpr prior to or coincident with ECCS actuation, thereby assuring cifective emaroency core cooling, include in your response what assurance is providea that this level measure.-nent will maintain its accuracy during blo.iown.
7.
Provide a list of all safety related setcoint settings used for rcactor opera tion.
For each setting tabulate (1) the setpoints uscd, (2) thn Technicai specification limits (or range), (3) the maximuu and/or.inimun value allcued (as derived from the safety analysis) prior te onset of 257.029
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conditions that are unacceptable from the standpoint of plant safety.
(i.es, safety limit)
The listing should clearly d1 fine the relationship between the Technical specification limits, the setpoint settings, and the maximum safety limit
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assumed in the safety analysis, and demonstrate that adequate margins are provided that insure actuation in a timely manner.
The listing should include the following, in addition to the rcactor protection and E5FAS setpoints:
(1) Core Flooding Tank Isolation Valve interlock setpoints.
(2)
Decay Heat Isolation Valve intericch setpoints. ',,
(3)
Auxiliary Feeduct2r Actu,a* ion,parc:eters (i.e., level and pressure setpoints).
l'.ain Steam Line Isolation actuation setpoints (i.e., pressure).
(4)
(5) Setpoints useci in change ovtr from injection to recirculation made of operation.
(6) Other safety related interlock setpoints.
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