ML19221B151
| ML19221B151 | |
| Person / Time | |
|---|---|
| Issue date: | 03/31/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| 15.3.4, NUREG-75-087, NUREG-75-087-15.3.3, NUREG-75-87, NUREG-75-87 15.3.3, SRP-15.03.03, SRP-SRP-15.03.03, NUDOCS 7907120531 | |
| Download: ML19221B151 (6) | |
Text
NUREG 75/087
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STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION SECTION 15.3.3 REACTOR COOLANT FUMP ROTOR SEIZURE AND REACTOR COCLANT 15.3.4 PUvP SMAFT BRE AK REVIEW PESPONSIBILITIES Primary - Peactor Systems Brancn (RSB) becondary - Accident Anal r is Branch (AAB)
Core Perforrance Branch (CFB)
Analysis [ ranch (AB)
I.
APEAS OF PEVIEW The events postulated are an instantareous seizure of the rotor cr break of the shaf t of a reactor coolant pu p in a pressurized water reactor (PWR) or recirculation purp in a boiling water reactor (BWR). Flow through the affected loop is rapid'y reduced, leading to a reactor and turbine trip. The sudden decrease in core coolant flcw while the reactor is at power results in a degradation of core heat transfer which could result in fuel dar. age.
Tre initial rate of reduction cf coolant ficw is greater for the rotor seizure event. However, the shaf t break event per"its a greater reverse flow through the affected Icop later during the transient and, therefore, results in a lower core flow rate at that t're.
This SRP aectior is :ntended to cover both of these infrequent transients.
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The review is concerned with the postulatad initial and long-tern core and reactor condi-l tions that are pertinent to the rotor seizure or broLen shaf t events, the nethods of thermal and hydraulic analysis, tre postulated sequence of events including tire delays prior to and af ter protective systen actuntion, the assu ed reactions of reactor systen components, the functional and operational charactcristic cf the reactor protection system in terrs of how it affects the seq w ce of events, md all cperator actions required to secure and raintcin the reactor in a safe cndition.
The resuits of the applicant's analyses are reviewed to ensure that values of pertinent system para eters are vithin expected ranges for the type and class of reactor under review.
The parareters include: reak clad te peratrre, p3ak fuel temperature, core flow and flow distribution (includ'ng hydraulic instabilities), chemel heat flux (average and hot),
minim;m critical heat flux ratio or critical power ratic. departure from nucleate boiling ratio,.essel water level, therral power, "essel pressure, stean line pressure (BWR),
steam line flow (BW') ad feedwat r flow (BUR).
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USNRC STANDAR7 REVIEW PLAN
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Rev. 1
The sequence of events described in the SAR is reviewed by t;oth RSB and ICSB. The PSB reviewer concentrates on the n(._d for the reactor protection systen, the engirieered safety systers, and operator action to secure and raintain the reactor in a safe condition. T her ICSB reviewer, as described in SRP Sections 7.2 and 7.3, concentrates on the instrurenta-tion and controls aspects of the sequence ccscribed in the SAR and evaluates whether the reactor and plant protectian and safeguards cor.trols and instrumentation syste s will function as assured in the safety analysis with regdrd to aJtoratic actuation, re"ote sensing, indication, control, and interlocks with auxiliary or shared systers ICSB also evaluates potential bypass rodes and the possibility of ranual control by the operator.
The analytical methods are reviened by RSB to ascertain whether the ratheratical rodeling and corputer codes have i;een previously reviewed and accepted by the staf f.
If a referer.ced analytical rethod has not been previously reviewed, the reviewer requests initiation of a generic evaluaticn of tne new aralftical model by AB.
AB, a, descrit'ed in SRP Stcticn 4.4, perfccms generic reviews of the ther al-hydraulic computer nodels used for tnis trar;ient.
AB also performs, u;.cn request, adjitional analyses related to these accidents for selected reactor types.
The salue" of all paraTeters used in a new analytical rodel, including the initial con-ditiens of the core ard system, are reviewed. It is the resconsibility of the R$B engineer to ccntact his counterpart in CFB to ensure that the relevant physic, data have been used in any ;taf f calculations.
AAB is notified regarding the extent of fuel f ailures that are predicted by the analysis.
AAB then evaluates the radiolo]ical consequences.
II.
ACCEPTANCE CRITERIA 1.
Tne basic objectives of the revies of the transients resulting f rom a rotor seizure at sraft break in a reactor coolant purp are:
To identify which of these infrequent transients is the more iiniting, a.
b.
To verify th3t, for the Fore limiting transient, the plant responds in such a way that the criteria regarding fuel damage, radiological consequen es, and systen w
pressure are r et.
2.
The specific criteria for the rotor seizure and shaf t break transients are:
a.
For everts such as the rotor seizure or shaf t break in a reactor coolant pump, tte plant should be designed to linit the release of radioactive naterial to assure that c":es to persons ffsite are kept to values which are 3 small frcction of 10 CFR Part 100 c- <elines.
b.
The potential for core damage should be evaluated on the basis of the acceM"'re c-iterion for DNBR in SRP Section 4.4 (Ref. 9).
It DNBR falls below the limits of this criterion, fuel damage (rod perforation) stould be assumed unless it Rev. I 15.3.3-2
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can be shown, based on an acceptable fuel damage rodel. that no fuel failure results. Any fuel danage calculated to occur rust be of sufficiently limited extent that the core will rerain in place and intact with no loss of core cooling capability.
c.
Pressure in the reactor coolant systen and main steam system should be rain-tained below llot of the design pressure (Pet. 1).
d.
A rotor seizure or shaf t break in a reactor coolant pump should not, by itself, generate a core sericus condition or result in a loss cf function of the reactor coolant system or containment barriers.
Only safety-grade equipment should te used to ritigate tue consequences of tte e.
accident. Safety functions should be accorplished assuming the worst single f ailure of a safety system active corponent (see Refs. 2 and 3).
f.
The ability to achieve long-tern coolabiiity of the core should be verified.
g.
This event should be analyzed assuming turbine trip and coincident loss of of f site power and coas1down of undamaged pumps, 3.
The applicant's analyses should be perf ormed using an acceptable analytical model.
The equaticos, sensiti vity studies, and rodels described in Pef er ences 4 through 8 are acceptable. If other analytical methods are proposed by the applicant, these rethods are evaluated by the staff for acceptability. For new generic rethods, the resiewer recuests an evaluation by fB.
The values of the para eters used in the analytical rodel should t'e suitably conscr-
- vative, The following values are considered acceptable f or use in the rodel, a.
The initial po.ver level is taken as the licensed core therr.al power for the neber !
of loops initially assured to be operating plus an allowance of 2. to account for pcwer reasurerent uncertainties, unless a lower power level can be justi fied by the applicant. The nurter of loops operating at the initiation of the event should correspond to the opcrating conditicn which rapinizes the conseq;ences of l
the event.
b.
Conservat' /e scrar characteristics are assured, i.e., f or a PWR r aximun tire delay with the rost reactive rod held out of the core, and for a BWR a design conservatist factor of 0.d tir es the calculat9d ne';ative reactivity insertion rate.
c.
The core burnuo is selecced to yield the mst liriting corbinatico of noderatcr temperature coefficient, void coefficient, Doppler coefficient, arial power pro-9 file, and radial power distribution.
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Pev. 1 15.3.3-3
4 For tnose art _- of review identified in subsection I c.? tnis SPP section as being the responsibility of other branches, the acceptance criteria and their "ethods of application are Contained in the SRP se tions corresponding to those branches I I I. E E.V_I EW PRO.C_E D_U.R E S The : rocedures t;elow are u<.ed during both the construction pern:t (CP) and operating license (OL) reviews. During the CP review, 'Ne voltaes of syste'- para"eters and,etpoints used in the analysis will te preliminary in nature and subject to change. At the OL review stage, final values sheald te used in the analysis, and the reviewer should compare chese to the limiting saf ety systw sett mgs included in the ;.r;;osed technical specifications.
'ne applicants' analyses of the rotor seizure and shaf t treak events are reviewod by PSB regarding the occurrences ltading to the initiating event.
The sequence of events, from snitiation ur.til a stabilized condition is reached, is reviewed to estertain.
1.
The extent to which n?rtally operating plant instrorentatico and controls are assu ed 'o function.
r 2.
The extent to which piant and reactor protection syster s are required te function.
3.
The creJ1t taken for the functioning of rarrally cperating plant systers.
4.
The operatior of engineered safety systm s that is required.
9 5.
The extent and time at which operator actions are roquired.
If the SAR states that one of the trarsiente is not as limiting as the other, the reviewer evaluates the justification pres <nted t,y the coplicant. The applicant is to present a quantitative analysis in the SAR of the transient that is deternined to be more limit;ng. For the transient that is found core limiting, the reviewer confirm that the effects of the transient arc deternined for each r1cde of operation (e.g., one, two,
three, or four-ioop) allo ed by the technical specifications. Eitner a separate ana!ysis should be presented for each rode of operation or the effects of eacn rode should be referenced to the limiting case.
For the more limiting transient, the PSB reviewer, with the aid of the ILSB rtviewer, reviews the tining of the initiatice of those protection, engirecred safety, and other systens needed to limit the consequences of the transient to acceptable levels. The R5B reviewer compares the predicted variation of system par)Pers with various trio and system initiation setpoints. The ICS3 reviewer evaluates automati: initiation, actua-tion delays, possible bypass codes, interlocks, and the feasibility of ranual operation if the SAR states that operator action is needed or expected.
To the extent oeemed necessary, the RSB reviewe.' evaluates the eifect of single ac tive failures of cafety systems and components which may alter the course of the transient.
15.3.3-4 nov 1 q
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This phase of the review uses the systen review procedures described in the SRP sections for Chapters 5, 6, 7 and 8 of the SAR.
The rathematical rnodels used by the apLlicant to evaluate core perfornance ar.J to predict system pressure ir the reactor cool 3nt syster and rain stea"1 lines are reviewed by R55 to determine if these r.odels have been nrevicusly reviewed and found acceptable by the staf f.
If not, AB is requested to initiate a generic review of the nodel proposed by the applicant.
The values of systetr parameters and ini+ial core and system conditions used s innut to tie model are reviewed by RSB. Of particular irportance are the reactivity coef ficients and control rod worths used in the applicant's analysis, and the variation cf rederator ter perature, void, and Doppler coef ficients of reactivity with core life. The justification
't ovided by the applicant tc show that he has selected t.c core burnup that yields the minirum ",argins is evalueted. CfB is tunsulted regarding the values of the reactivity pirareters uced in the applicant's analysis.
Tho results of the applicant's analysis are reviewed and compared to the acceptance cri-teria in subsection !! regarding the ra<irum pressure in the reactor coolant and rain st m systens. The variatiuns with tire during the transient of tne neutron power, heat flumes (average anJ n axinun), reactor coolant syste" pressure,iJnirun DNBR (PWR) or CPR (LWR), core and recirculation lcop coolaat flow rates (SWR), coclant conditions (inlet temperature, core average ter rerature (PnR), core average steem volt.ae f raction (BWR),
average e xit and hot channel exit ter peratures, and steam fractio 15), te perature (naxi"um f uel c enterlina te"perature, r axinun clad temper at;re, or maxirum f Jel enth31py), stea+
line pressure, cor tainrent pressure, pressure relief valve flow rate, au flow rate f ron the re> _ tor cool ant sys te": to the containnent system (if applicable) mre reviewed. The rore it orti nt of trese carareters (as listed in section I of tnis W ) are compared to those t redir.ted for < r s milar plants to confirn that t6 ej are within expectcd range. In particu;ar, the reak cladding tm perature and percentage of fuel rods that experience a departure from nucleate boiling condition are compared and AAB is notified regarding tre extent of fuel f ailures predicted ' y the analysis.
a IV.
EVALUAT KN FIND N S
'he revieaer verifies that the SAR contains suificient information and his review supccrts tne follouinc L,inds o' statements and ccrclusions, which should to included in the stat.'s safety evilnation re;crt:
' The analyse, arc ef fet ts of an instantantous seizure of a rotor and an instantaneous brNL of a shaf t of a reactor coolant pu"p* dur%g any allowed rcde of operation have been reviewed. It was found that the r. ore lini'ing of these events was the Nis event was evaluated by the applicant using a ratteratical model that had been previously r eviewed and found acceptable by the staf f.
The parameters used as ir.put to this rodel were reviewed and found ; be suitaHy con-sersative. The results of the anvlysis showed that _ _ of the fuel ruC axperienced
- Recirculation puno shaf t f or a EWR.
15.3.3-5 Rev. 1 kr
departure fron nucleate boiling (CNB) and that the peak clad temperature reachtd was if. This assures tN; the f uel damage will be ninic al and that no loss of cr re cooling capability will result. The analysis showed that the raxitm pressure within the reactor coolant and main steam s; stas did not exce: d 110 of the design pressures "Thc staf f concludes that the plant design is acceptable with reg 1rd to a russible seizure of a rotor or break of a shaft of a reactor coolant pump.
REFEFENCES 1.
ASME Boiler and Fressure Vessel Code,Section III, Nuclear Power Plant Corponents,'
Article NS-T00, " Protection Arpinst Overpressure, ' A erican Society of Mechanical E r,1 r e e r s.
2
" Staff Discussion of Fifteen Technical Issues Listed in Attachrent to Nove-t'cr 3, 1976 Merorandum fron the Director, NRR, to NRR Staff" (Issue No. 1), U.S. Nuclear Regu!1 tory Com,ission, NUPEG-0133, Nover ber 1976.
3.
"Staf f Discussion of Twelve Additional Technical Issues Raised by Pesponses to November 3, 1976 Me orandum from the Director, NRP. to NPR Staff" (Issue No. 22),
U.S. Nuclear Regulatory Cornission, NUPEG-0158, Decer ber 1976.
4 F. M. Bordelon, " Calc'ilation of Flow Coastdown after Loss of Reactor Coolant Purp, WCAP-7973, Westinghouse Electric Corporation, August 1970.
S.
C. D. Margan, H.
C. Cheatwood, anJ J.
P. Glanderm ans, "PADAR - Reactor TFerral and Hydraulic Analysis During Peactor Flon Coastdcwn,' BAW-lC069, Babcock and Wilcox Corpany, July 1973.
6.
R. H. Stoudt and J. E. Busby, "CADD - Computer Applications to Direct Sir'ulation of Transient Events on Water Reactors," BAW-10080 (r.coproprietary) and BAa'-10076 (proprietary), Babcock and Wilcox Conpany, July 1 R 3.
7.
'Syster 80 Standard Safety Analysis Peport (CESSAR), Combustion Engineering, Inc.,
A; gust 1973, a,
P. Linford, " Analytical Petnods cf Tran' 'cnt Evaluations in the General Electric Boiling Water Reactor," NEDP-10802, General E?ectric Co"pany, April 1973.
9.
Standard Peview Plan 4.4, " Thermal and Hjdraulic Design.
O Rev. I 15.3.3-6 1
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