ML19220A799
| ML19220A799 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/23/1975 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | Moore V Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904240625 | |
| Download: ML19220A799 (8) | |
Text
{{#Wiki_filter:. h;w _-O ocke:.30. : 30-320 Y. A. Macre, Assistant Director for LaRs, Srcup 2, ilt SECCiiD ROU.iD QUESTI0iis, THREE MILE ISLRiD ;iUC' EAR PLNsT, U:iIT 2 Three Mile Island 2 Pl ant !ia'ne: Docket Nu:rcer: 50-320 Licensing Stage: Operating License Responsible Branen LWR 2-2 and Project ;.eader: B. Washourn Technical Review 3 ranch Involved: EIaCS Branca Description of Review: Second Round Questions Requested Ccepletion Dato: April 11,1975 Review Status: Awaiting Information The enclosed list of questions was prepared oy tne tiRA:RS Electrical. Instrumentation and Control Systems Branch. Inase questicns and positions are the result of our evaluation of :ne F5AR :nrougn Amendmnt 26. We have evaluated the responses obtained to our "i"' set of . c in tne questions and concluce that concerns continue t folicwing areas. (1) Documentation relating to seismic and environmntal qualification of safety-relatad equipmnt. (2) Testability of safety related equipent. (3) Post-accident inonitoring instrumntation. (4) The auxiliary feedwater system design. (5) Concerns relating to the Diesel Generators. (c) Updating tne figures and drasings in t he FSAR to reflect the modified designs. (7) Reco:nrendations concerning the Technical Specifications. It seculd be no*ed that the applicant has not res; ended to the part of Item 22.11 (wnicn addressa tne degree of con 'ormance of the design to Regulatory Guide 1.47) of our first set of questions. However, it is / 7 9 0 4 2 4 9gg-P o.
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.s,;.;: 3. m Victor Stallo, Jr., Assistant Director for Reactor Safe *g/ Division of Technical Review Office of Nuclear Reactor Regulation Encicsure: Second Sat of Questions CC* S. Hanauer F. Scnroeder A. Gia::russo K. Kniel
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wa a u 22.25 Your respcnse to Item 22.1 is not complete in that the (3.10) results of seismic qualification for several items have not been p: cvided (a note in Table 22.1-1 indicates that this will be provided later). Provide tnis missing i n fo rma ti on. Also, include a statement fcr each item of equipment tc the effect tnat the testing and/or anaijsis performed quali fied tne equipment for Category 1 service as proposed for use in :nis plant. 22.27 Supplement the respcnse to Item 22.2 by providing cespcnses (3.11) to the folicwing: 1. Table 3.11-1 of the FSAR does not indicate the limiting CSA ccnditicns in the containment with regard to the chemical envi ronmen t. Provide this information. 2. Justify not qualifying. ne instrument cables associated with the reactor coolant pressure instrumer.ts for the acci dent envircnment. Addi ticnally, provide succc rting information for tne statement that a failure in tnese cables will generate a fail safe signal for ne SFAS. 22.2S Identify all safety-related equiccant or components that (7.0,3.0) will not be tested during normal ;cwer operaticn of the reactor. Also, for each piece of equicment or ccmconents identified, specify and ;ustify tne ir terval between tests for that equipment cr ccm;cnent. 22.29 Cur positicn witn regard to protecticn systems response (RSP) time testing is that: ( 7. 0,16. 0) 1. Periodic tests for verification cf system respense ti es of the Reactor ?rctection Sys ter and the Engineered Safety Feature Actuation Systems sncuic include tne rescanse times of the sensors whenever practical, 2. In some cases, indirect means cf verifying sensor response times may be usec. Catails of sucn indirect means of verifying sensor resconse times snould be included in acplications and will be reviewed by One staff. 3. Excepticns to the abova should be specifically ider.tified and justified. Y0-008
=-- 22-2 We will require conformance with this position. Therefore, revise your Technical Specifications accordingly. 22.30 For each of the valves identified in tne response to item ( 7. 3.1.1 ) 22.8 provide tne folicwing. 1. Verify that the air storage reservcirs for tnese S C' es have been seismically qualifiea. 2. Describe any integrated system test (i.e., e test which includes a simulated SFAS, sciencid pilot valve ccerators, air storage reservoirs and actual valve closure).unich will be conducted periodically fcr eacn valve. 22.21 The response to Item 22.10 states that procedures have been (none) reviewed to ensure that the removal of a redundant cart of a system must be acccmoanied by testing of tne remaining portien o,ior to or immeciately af ter cisconnecting the portion af fectec. Provide tne basis for choosing to test either prior to or af ter disconnecting tne affected subsystem. 22.32 Our positicn with regard to post-accident monitoring and (RSP) safe shutdcwn instrumentation is that tnese instrumentation (7.5.1, / - systems should be: .3.o,r 1. Redundant with indicators in the centrol ream for bcth channels with at leas one channel reco rded. (The recording system, recarters and associated circuitry and ccmponents, are required to be seismically qualified to verify tneir cperability folicwing, not necessarily during, seismic events). 2. Energized frcm tne ensite Class Ir pcwer su; plies. The information contained in FSA? Sections 5.1, 7.5.2, Table 7.5-1 and the response to Iter 22.12 concerning post-accident monitoring and safe shutcown instrumentation does not clearly indicate conformance to the above. Therefore, verify compliance to this position er provide justificaticn for any exceotions. 22.33 It does not acpear trem the rescense to Item 22.1? that a (ncne) complete review of your operating, aintenance inc testing procedures was conducted to determine tne extent o'f usage of jumpers or other temporarv forms of oypassing fur.cticr.s. t 7C-009
22-3 for operating, test:ng or maintenance or safety related systems completa Inis review and provide the requested information. 22.34 Your response to Iter 22.15 is inacequate for an evaluaticn (6.0, 7.0, of the Emergency Feedwater (EF) System design. Provide 15.0) a Failure Made and Effects Analysis
- .o demcnstrate tne operation of at least one EF subsys tem, i.e., either the motor driven or turcine driven sucsysten, under the follcwing conditions, considered one at a time:
1. Loss of offsite pcwer concurrent with loss of all the standby diesels. 2. Loss of any battery, one at a time. 3. Loss of any Vi al A-C Control bus, cr.e at a time. 4. The loss of pressure, or overpressure, cn any control air system required (if any) for operation of tre EF system. 22.35 Provide restceses to the following with regard to the Ciasel ( 8. 3.1.1 ) Gen e ra to r s. 1. Secticn 3.3.1.1.8 of the FSAR s tates that botn d'ecel-engines will ce au:cmatically star:ed by a SFAS signal or ucon the occurrer.ce of Icss of vol: age of either engineerec safety features bus. Ho..ever, Section
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incicates :na: a permissive for the fully autc atic star: made of oceration of tre diesei engine is less of the 2160 V inccming sucply f-cm either of tna two auxiliary transformers to tne engineered safety features cuses. Resolve tnis apparent inconsistency. 2. The response to Item 22.18 cces not indicate tnat alarms will annunciate in tne main control rcem i f the shutdown pushout cm located on :ne generating uni skid is not reset. Accordingly, veri fy that alarms are provided for tnis condit cn. Also, indicate for this ccndition if the fully au cmatic start Tcde of operaticn for One ciesel engines may be reinstated from the main cc trol room. e.b,()j[0)
,a 9 -,- 3. Confirm that the Diesel Generator Sets for Three Mile Island Unit 2 are icentical to tncse used in a praviously licensea nuclear piant as indicated in :ne response to Item 22.3C and icenti fy tne plant. Identify and justify any differences in tne design of your diesels. 22.36 Figures 8.3-3, 3.2-5 and e.3-7 shoulc be revised as stated (3.3) in tne response to Item 22.24 22.37 Table 2.3-1 of the itnhnical Speci fications should be (16.0) revised to indicate ali Reactor ?rctection System trip setting limits. Of prima.s concern in nis regarc is One trip setting limits of tce vc'tage and current relays for tne reactor coolant pumo pcwer moni tor trip. 22.33 Recent operating ex;erience with P'.;R's has indicated tnat (none) single failures in the centrol system nave resulted in uncontrolled withdrawals (rot ejection) of a single rod cluster control assembly (RCCA) from the core. As a resul t of these cccurrences, v.e reques t that you cerferm an analysis to determine that tne consecuences of uncontrolled withd- 'l of a single RCCA under any cassible ccnditions of re- _ c operaticn do not resuit in exceeding specified acceptable fuel design limi ts. If tne results of this analysis shcw nat these limits are exceeded, you must provide tne results of a Failure Mode and Effects Analysis to shew that a single failure occurring in tne control system will not cause tne uncentrolled aitndrawal of siigle RCCAs. If tne resalts of :nese analjses shcw that it is possible for uncontrailed withdra.sais of si gle RCCAs to c.ccur and that specified acceptable fuei desi:n limits coult be exceeced as a result, then your proteciicn system must be designed to cetect and terminata t e resulting transient before exceeding the fuel design limits. Perform the analyses anc provide tne informaticn as indicated alive. 22.33 it is not clear from your accident analysis what the (7.0,15.0) r+ uirements are regarcing disconnection of tne Reacter Cc.iant.cumos from their pcwer scurces (because of degraded frecuency or fsalt conditicns of tre acwer sources) to alla.v for full coastdcwn of One pumps. Therefore, provide tne following addi ticnal information. '/O- 011 -e
22-5 1. In the event that full cumo coas tdc.<n is needed to prevent exceeding core safety limits, Orovice an analysis to describe tne effec:s on Cumo coac tccwn capacility in the event tnat ne pump breakers failed to isola te the pcwar surcly during an underfrequency con di ti on. Your response shcuid define the limiting underfrecuency condition including the maximun allcwable frequer.cy cecay rate. 2. If credit is taken for reactor coolant pumo coastdown in tne accident analyses and the pumos must be dis-connected On grid underfrecuency condi tions, the cumo breakers and associated uncerfrecuency trips must be cesigned and qualifiec in accordance with the require-cents of IEEE Std 279-1971 and IEEE Std 30E-1971. Describe the degree of conformance of your design witn this posi tion and j us ti fy any exceo tions. O O 'iG 01g =}}