ML20155J316

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Insufficient CSP Npsh
ML20155J316
Person / Time
Site: Fort Calhoun Omaha Public Power District icon.png
Issue date: 06/15/1992
From: Ruhl G, Van Sant B
OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20155J310 List:
References
CID920473-02, CID920473-2, NUDOCS 9811120103
Download: ML20155J316 (12)
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{{#Wiki_filter:. FORT CALHOUN STATION FC-154 GENERAL FORM R13 l NUCLEAR SAFETY EVALUATION Reference N0D-QP-3 ID No. d/b 920473/es SECTION A Page I of IO (from 9.1) 10 CFR 50.59 Applicability Screening 9.1 Activity Identification Procedure Change No. A/A affecting Procedure dA l Modification Request No. MA Design >4 Installation [ ] Testing [ ] l Temporary Modification No. //A Engineering Change Notice No. #A Other CIb 9Lo173l02. Document

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14 enc 1. Nuclear Safety Evaluation Conclusion [] This activity is not a 10 CFR 50.59 activity, because it: Does not change the facility as described in the USAR. Does not change procedures as described in the USAR. Does not involve conducting tests or experiments not described in the USAR. Does not affect Nuclear Safety in a way not previously evaluated in USAR. N This activity is being done pursuant to 10 CFR 50.59. This safety evaluation must be reviewed by SARC; ref Tech. Spec. 5.5.2.7. This activity must be reported in the annual report; ref 10 CFR 50.59, Item b, Paragraph 2. [] This activity involves an Unreviewed Safety Question. The activity must be I canceled, or revised and re-evaluated, or NRC authorization is required prior to implementation; ref 10 CFR 50.59, Item c. l We hereby certify that this Nuclear Safety Evaluation is complete and accurate to the best of our knowledge. I B_g g aqg Prepared by ba/V D kukI Usa Resten //#-/92-Time /d/E Date lo E 2 Signature p pq2 jg /$. 0. fA >J AMJ Date 6 //s'bf 2 Time l6:.3 0 Reviewed by / Print Na e I N atA Extension 2 97 ~7 Signature A-1 l 9811120103 981029 f5 PDR ADOCK 05000285L, P PDR

? FORT CALHOUN STATION FC-154 GENERAL FORM R13 NUCLEAR SAFETY EVALUATION l Reference N0D-QP-3 IDNo.C/D920f15/oZ SECTION A Page

2. of fo (from 9.1) 10 CFR 50.59 Applicability Screening 9.2 What (specifically) is being done?

I4SA$ See%n 6 is beiny af,+ed -/o revise 1%e def /%silive %efion ll erd d (MPstl) calcaldio ts L. +4a recircu la fion made in a cc o rdent e wi+h reced revis ion +e -Ile plan & desty Lasis fs refleef as I.all+ c.andi+ ion s. This in cluele s credi+ing -14 e a vailable Nfsil wifb sump sulecolty herd u>he<e previsasly +4 ;s cred ; + wrs nof alla ae.d. 9.3-Why is this being done (briefly)? LFA 12-016 includes to e<e cRve aebian -fo updsfe de usAR anal sis resulft whick k.se d as Lu;/+ hyderalie y em In d i e r+e. that au;la ble Nfsil e,Jeal,+;sa.s fo< +4e \\ recirtu.laNbn thode of confrinMenf sprs.y (CS) reg uire crediking sump sulcooliny herd +a me e-I-d e-f u.mp (^egu ire d N/ 5 //. 9.4 Does the activity involve a change to the Technical Specifications? PQ NO This activity meets the requirements of current Technical Specifications. The following sections were reviewed: 2.3i 2,4 Continue with 9.5 ['] YES - Technical Specification Section must be revised prior to performing this activity, l Exit this procedure and continue with N0D-QP-7. A-2 FC/ FORMS

l FORT CALHOUN STATION FC-154 -GENERAL FORM R13 NUCLEAR SAFETY EVALUATION l Reference N00-QP-3 ID No. C/b 92 0413/07-SECTION'A Page 3 of_/0 (from 9.1) 10 CFR 50.59 Applicability Screening 9.5 Does the activity involve a change in the facility? ' N NO Go to 9.6 [ ] YES Is this aspect of the facility described in the USAR? List USAR Sections reviewed: [ ] NO Go to 9.6 [ ] YES list USAR Sections Does the USAR description require any changes or revisions due to this activity? []NO continue with 9.6 10 CFR 50.59 applies to this f [ ] YES activity { Section B of the Nuclear Safety Evaluation must also be completed. Continue with 9.6 9.6 Does the activity involve changes to procedures? hd NO Go to 9.7 [] YES Are related procedures (including definitions or descriptions of activities or controls over functions).SAR? outlined, summarized, completely described, or implied in the U List USAR Sections reviewed: []NO Go to 9.7 [ ] YES - list USAR Sections Does the USAR description require any changes or revisions due to this activity? []NO - Continue with 9.7 [ ] YES - 10 CFR 50.59 applies to this activity l Section B of the Nuclear Safety Evaluation must also be completed Continue with 9.7 l A-3 FC/ FORMS-

FORT CALHOUN STATION FC-154 GENERAL FORM R13 NUCLEAR SAFETY EVALUATION Reference N00-QP-3 ID No. c/b 91 973 /o z. SECTION A Page 4 of /6 (from 9.1) 10 CFR 50.59 Applicability Screening 9.7 Does the activity involve tests or experiments? bd NO Go to 9.8 [ ] YES - Is the test / experiment one which.has been previously anticipated in the USAR? l [ ] YES list USAR Sections Go to 9.8 []NO - (i.e., it is not described in the USAR; including one-of-a-kind tests or new system configurations) Could this test / experiment degrade the margins of safety during normal operations or anticipated transients, or could it degrade the adequacy of structures, systems or components to prevent accidents or mitigate accident conditions? [ ] NO - Continue with 9.8 ( ) YES - 10 CFR 50.59 applies to this activity Section B of the Nuclear Safety Evaluation must also be completed. Continue with 9.8 .A-4 FC/ FORMS-

FORT CALHOUN STATION FC-154 l GENERAL FORM R13 i NUCLEAR SAFETY EVALUATION l l Reference N0D-QP-3 ID No. C ab 92.o */73/o 2. SECTION A Page f of /6 (from 9.1) 10 CFR 50.59 Applicability Screening ) 9.8 Could the activity adversely affect nuclear safety? Explain [ ] NO i Go to Nuclear Safety Evaluation Conclusion or continue with Section B of the Nuclear Safety Evaluation, if required. >:3 YES - How Revisien do -llte UMA N/Sil celeuldies could resulf in a desian basis sd +U -Me assoeMed cs peo s lo nd W e. ade m he aallal,le NP.sil. 'wi+k in s u ffteiend M/sil eNilcJte % omo s could ev i +s+e and f,il premheelv 5 deliNr (buired floa. &;/are ' mould be' less op cs pes + K4.5 ~ Word cae sta taald t4ae+ Ke-conhinmen& pexk prenure. Has this effect been previously evaluated in the USAR? discussed in USAR Section [ ] YES L Continue with Nuclear Safety Evaluation Conclusion 4 NO 10 CFR 50.59 applies to this activity Continue with Section B of the Nuclear Safety Evaluation A-5 FC/ FORMS l

FORT CALHOUN STATION FC-154 GENERAL FORM R13 NUCLEAR SAFETY EVALUATION Reference NOD-QP-3 ID No. d'DTE8'I73/01-Page _[;t_ of /0 (from 9.1) SECTION B Unreviewed Safety Question Determination 10.1.1 Identify Plant Specific Design, Operating and Technical Documents Document Title ID Number Revision Co,,hinne d 5yrny bab gbs p-c s-I3 l R3 usAA % i m e 3 S e.e + io n 6 AG L4r,ic,1 Sncifiedkt Seek 2.4 $,$'ats ,$o L e Ins, 5 - cal c - o z. I bD 10.1.2 Identify Applicable NRC Documents / Industry Standards Title ID Number Revision AEC Safely Lide /. k$ 0 .1.3 Identify Related Drawings Title ID Number Revision GE+CS 4Eb f*2.5 7C L - 210 - 13 0 55 sr-aA,8 A Pamp d-cie.s Carfrid,oe n Frme. 03c e l 10.2 List safety functions the affected structureg or ~7A c M euws N/ corr 7alo rd e nd comp +onents perform: pos L OC H -fo est s are. smk s w',-e la to nfm twent I ' h en no+ exceed t,n o's ? La c u s s /sh a, u>aie r-frem -tt e s ru r ud Doc + #A S &sL -tt!/ G sWA inmers+.Su n.s. W I List applicable accidents for which these safety functions are required: 60CA 4 B-1 l l [ FC/ FORMS. t

FORT CALHOUN STATION FC-154 GENERAL FORM-R13 NUCLEAR SAFETY EVALUATION l Reference NOD-QP-3 ID No. dlD 710'OND'l-Page 7 of IO t (from 9.1) SECTION B Unreviewed Safety Question Determination l-10.3 System Interactions Analyses [. Criteria ADolicable Criteria ADolicable Fire Protection ( ) Structural Impact .( ) Electrical Equipment Separation Criteria '( ) Qualifications ( ) High Energy Line Break -Review ( ) Possibility of Operator l . Seismic Interaction and Qualification ( ) Heavy Loads (.) 1 Electrical Systems Analysis [ ] Impact on HVAC [ ). Human Factors Review [ ] System / Component Security Review ( ) Natural Phenomena ( ) Environmental Radiological Release [ ] Installation of Temporary Modifications ( ) i Materials compatibility-( ) l Testing of Temporary l ' Containment Integrity b<:1 Modifications [ -)

. control Room Habitability

[ } Other: ( ) Missile Protection ( ) I Discussion of Applicable Systems Interactions Analyses y l' (Include Attachment Sheet as needed) //o s kys it / or e/cre Nm l cha.ste e) s.r e involve d w iS $t. D rop s &ckth e cr dikrt.9 e sul e.o l tu M de retiredbw uaclC M/bd milalle ca l$ls.ks ets. kalemle MfS// is 90a s'id le lutaler % reu des t% L,utr did cre.J [fs on/v 25% of 1h a v4 /d /c s w 3 &,."IN a lexd. fyiA/c y.senf peeb B-2 g gg, d 36,. S y s }<m s kQe%, i FC/ FORM, M;Y * .-m,_,---

FORT CALHOUN STATION FC-154 GENERAL FORM R13 NUCLEAR SAFETY EVALUATION l Reference NOD-QP-3 ID No. c;b 92.oV7 3 /c L Page b of JO (from 9.1) SECTION B Unreviewed Safety Question Determination 10.4 Could the proposed activity increase the probability YES [ ] of occurrence of an accident previously evaluated in the USAR? NO M Explain: Yo Cbt<n9e1 blIn ci M%de ko $e eX ISSin 9 <Ee 5 Y=,n

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I J v n,peAhk of ev. sis m en -l-u>$I CL Lo wl d increxse Ye a bic> b a ls ' lt hv bh o c cu.eren te A an o_ cc. < b. e nh. I / 10.5 Could the proposed activity increase the consequences YES [ ] of an accident previously evaluated in the USAR? NO b<Q Explain: $e fevis Eon Y 'lbe CLSINn betis c1 Vat'l< l {e t/fS// [u.- ~f{xe. CC p i [#s l'ec ?re u {c 'l ten ode cloes ncd c< Ner -M e consewen ee s of eu, nee IclenI se' n e acte m 4e. ANSil is sla + V -h, (re. edId[e. .S v s tem Merections wedisNsd in 5 4 0 W o 2.. ( b ee si+u k e.1 shee4 H of fjo 3 Could the proposed activity increase tfi'e}t. 9,bability fjpp' YES).ph %oSW 10.6 of occurrence of a malfunction of equipment important7 PF@.[/5T' pro ] to safety previously evaluated in the USAR? NO >,-e. E orofalll3v f oferar*Pmedh is1}orfqnY hb S4 )( ) ch O CL Wr f t 6Lt O A In AL o rt r b not irt teerted. l 10.7 Could the proposed activity increase the consequences YES [ ] of a malfunction of equipment important to safety previously evaluated in the USAR?- NO K Explain: be conter sentes c$ a wtd0ntNon & en we'pnten h in19erU 4, ss.k is n 4 e ffe d ect. 7Le es suas

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i clen%ed -is s w iv > 2000 C,PM socav oo si - MS sin te eduak J / I Altsil rr stw - -h> be adM te, $r allmoies%.s% epecaHen m s te s B - Q ;, n y l. " a on w&Inun+ -tr FC/ FORMS

FORT CALHOUN STATION FC-154 GENERAL FORM R13 NUCLEAR SAFETY EVALUATION Reference NOD-QP-3 ID No. 6/b 9 2.o'/*?.3 /ot. Page 9 of_10 (from 9.1) SECTION B Unreviewed Safety Question Determination .10.8 Could the proposed activity create the possibility of YES ( ) an accident of a different type than any previously evaluated in the USAR? NO (><Q Explain: 1lte CS p~os All se< fem Keir su $ehy te luhel kneken 8 I / as evaluakt.A h -L urAR. Alo shnl 9Lvs % l er optrdioital l ' i eknases cill Le ntede and -)dere Are 4L a-se ss ibilik sf J l i a rt etec id en h oh a dih$ered -[v d e i5 ILo$ Crerh). /I 10.9 Could the proposed activity create the possibility of YES ( ) a malfunction of equipment important to safety of a different type than any previously evaluated in the NO pq USAR? Explain: Msl[undion ob d u e. " b Ist R d E 9 vts E R e CS hss 1 V t/f51-{ 1s nof e x /w led to fle. t)s42. 1'L e ele s,'n n basis mis ton indicks L4 adede Nes ti ts o~ildie J credH Ie.ul bued o f S w L c o olitt, c.,c{nininenf hsnsierth arial,vsis t o ro dh. 1%crehre. N tossikIMy of a malGenefion bf eu.hed I.,wh4 -fo s afe.b of s. cli FR.re A+ +ype 'cauld na+ b e-c re u kel. 10.10 Does the proposed activity reduce the margin of safety YES [ ] as defined in the basis for any Technical Specification? NO p<3 Explain: 7(e LIS48. Section (2 2-l and A60 Ss[cbv [ wide. I c[n ne/ de[ine - b ka.t is of a F kcqin of b eis. I c Sde4v ludie is su Le.ee, W, ca s~,,ebea se.Ldel sv, (-<celut in, AfsH aNildle. L orteimi a. sa .i i s tHk/L wt.lvsir Cllawr -tLis ne-L ds lasv - Irwe ve.c, i f is accepkble k ai CA C44 A. Ah A e Y b e / Me-0 3-e e Ml(A p tC. C.u On ren.W e ~ n +1 % -fla od n.os deri .s l% openLilNu dni da as+ l B-4 ivts<pn eP safeh $r Ke_ mi/d/c #FS//. S't>ne adep., k o.4dre rs A I# ^^ "/ N'

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FORT CALHOUN STATION FC-154 GENERAL FORM R13 NUCLEAR SAFETY EVALUATION Reference NOD-QP-3 ID No. 6 / b 9 2.o '/ 73 /o 2. Page /0 of_/0 (from 9.1) SECTION B Unreviewed Safety Question Determination 10.11 Summarize USAR changes which are needed p_r attach marked-up copy of affected pages: See afbAed m e k.e.ct - a f a.3 c s. 10.12 Annual report of 10 CFR 50.59 changes, tests and experiments. Provide a brief description of the activity: bis ctc.fWify IA volves revis ieth SL be5 (qn bcd it cA lt u. I Nc> n o$ i J J milJ le M t s FI />< +4 e sr p-is rn ret irtwla% m o d e.. Nir IncIk1cf b.o d d of LISAd h Ic[-/ecf h/e e. t r1ca Aede L uis +o in c \\ u c( e u, sulc oa lcm he d cre.L:+ & < a n s h a^ilulle Al t s H. Summarize the safety evaluation: ~I/4is acfivib [5 [t/4s 2*ne / J nrsurn4 h lO CPA 50.5'l bd does tLoh dears.be Nue lene Subhf s a lae.e n.a a -de a vaiIdle NPSll b n ed a-m.a m o'n of s~fek 4.- a s -f(e Tec4nlc-l 5 e. eih e<41 e n t u d Wie f vsicsI anel onco /AnQ 4 7 l cd p e-c.h of Me S VIMS Are n e d f, [ [ e_e. k cl. Go to the Nuclear Safety Evaluation Conclusion B-5 FC/ FORMS

&Ib 9t0413l0 2. ~ kleby Erin % FORT CALHOUN STATION next FC-154 GENERAL FORM R12 NUCLEAR SAFETY EVALUATION Reference NOD-QP-3 ID No. Idg-92 d 3 Page // /[ of (from 9.1) ATTACHMENT SHEET

4. W No physical deficiencies are present as a result of not meeting the

~ requirements of the current licensing basis to use AEC criteria Safety Guide 1.1 for calculating pump NPSH. While normal M engineering practice allows for subcooling in calculating the NPSHa, the AEC criteria conservatively directed that this not -be credited to build in an inherent safety margin and eliminate the possibility of a inadequate suction head. Based on the available NPSH from subcooling it is apparent that in the event of a LOCA a significant margin for NPSHa exists by the use of actual sump temperatures (Ref. C-E letater 0-MPS-91-120 dated 8/23/91). The pumps are currently lacking less than three feet of static head and have available more than 20 feet of head from subcooling (Based on EA-FC-90-94). 3 ~~ m M M W m FC/ FORMS

9 11/2/70 ( (Reprinted 12/1/70) SAFETY GUIDE 1 NET POSITIVE SUCTION HEAD FOR l EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL SYSTEM PUMPS i A. Introduction possible accident conditions. For example / if proper operation of the emergency core cooling f i Proposed General Design Criterion 41 re. system depends upon maintaining the contain- ^ quires that the emergency cooling and contain-ment pressure above a specified minimum ment. heat removal systems be capable of ac, amount, then too low an internal pressure (re-complishing their required safety functions as-sulting from impaired containment integrity suming partial loss of installed capacity. In cur-or operation of the containment heat removal rent designs the ability to accomplish these systems at too high a rate) could significantly safety functions reliably depends in part on the affect the ability of this system to accomplish proper performance of system pumps which, its safety functions by causing pump cavita-in turn, depend.4 on the conditions under which tion. In addition, the deliberate continuation of the pumps must operate. One of these condi-a high containment pressure to mainta,n an i tions is suction pressure. This guide describes adequate pump NPSH would result in greater a suitable relationship. between increases in leakage of fission products from the contain-containment pressure caused by postulated loss ment and higher potential offsite doses under of coolant accidents and the net positive suc-accident conditions than would otherwise result. tion head (NPSH) of emergency core cooling Changes in NPSH for emergency core cool-and containment heat removal system pumps Ing and containment heat removal system which may be used to implement General De, pumps caused by increases in temperature of sign Criterion 41. the pumped fluid under loss of coolant accident B. Discussion conditions can be accommodated without rell-ance on the calculated increase in containment l A significant consideration related to emer-pressure. Adequate NPSH can be assured by gency core cooling and containment heat re. locating pumps at suitable elevations with re-moval systems is the potential for degraded spect to the storage volumes connected to their pump performance which could be caused by a suction sides, by using multistage or booster number of factors, including inadequate NPSH. pumps, by a combination of these methods, or If the NPSH available to a pump is not sum. D cient, envitation of the pumped fluid can occur. This cavitation may reduce significantly the C. Regulatory Position capability of the system to accomplish its safety Emergency core cooling and containment functions. heat removal systems should be designed so It in importalit that the proper performance that adequate net positive suction head (NPSH) of emergency core cooling and containment is provided to system pumps assuming maxi-heat removal systems be independent of calcu-mum expected temperatures of pumped fluids lated increases in containment pressure caused and no increase in containment pressure from L by postulated loss of coolant accidents in order that present prior to postulated loss of coolant l to assure reliable operation under a variety of accidents. 1.1 -}}

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