ML19221B148
| ML19221B148 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-15.2.8, NUREG-75-87, NUREG-75-87 15.2.8, SRP-15.02.08, SRP-SRP-15.02.08, NUDOCS 7907120525 | |
| Download: ML19221B148 (7) | |
Text
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STANDARD REVIEW PLAN
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OFFICE OF NUCLEAR REACTOR REGULATION SECTION 15.2.8 FEEDWATER SYSTEM PIPE BREAKS INSIDE AND OUTSIDE CONTAINPENT (PWR)
_R E V I E W R E S P CN_S I_B__I_L I T_I E_S Prirary - Reactor Systems Branch (RSB)
Secondiry - Accident Analysis Branch ( AAS) suxiliary and Power Conversion Systems Branch ( APCSS)
Containment Systers Brancn (CSS)
Core Perforrance Branch (CPB)
Electrical, Instrumentation and Control Systers Cranch (EICSB)
Pechanical Engineering Branch (PEB)
I.
U cAS OF REVIEW Tie transient th3t results fron a postulated feedwater line breek is e nsitive to the
. isch 3rge rate; consequently, a range of break sizes <hould be evaluated bCth inside and outside containrent to detemire the acceptability of the response. Cepending upon the size 3rd location of the brPak and the plant operating Conditions at the tire of the break, the break could cause either a reactcr coolant systen cooldown (by excessive energy discharge through the break) or a reactor ccolant systen heatup (by reducing feedwater flod to the affected stean generater). Therefore, analyses of various postulated break sizes and loca-tions are needed to identify the particular situatior. that is rest limiting with respect to system effects.
If a ftedwater lire rupture causes the water in the steam generator to be discharged through the break, the water will not be avail 0t;le for decay heat re oval af ter reactor scram. The break locatien and size ray be such to pre,ent additicn of any feedwater to the affected stean generator. An auxiliary feedwater systen is therefore provided to assure that feed-cter is available to provide decay heat removal.
The reviaw includes evalation of the applicant's postulated initial core and reactor
~.n-diticos pertinent to the feedwater line break, the rethods of therral and hy#aulic analysis, the postulated secuence of events including analyses to deterriine the ti e of reactor trip and tire delays prior and subsequent to ir.itiation of reactor protection systen actions, the assu~ed response of the reactor coolant and auxiliary systems, the functional and operational characteristics of the reactor protection sy, tem in ter~s of its effects on the sequence of events, and all cperator actinns required to secure and raintain the reactor in a safe shutdown condition. The results of the analyses are reviewed to ensure that the values of 79n71709 USNRC STANDARD REVIEW PLAN Sv.nd.,d,.. pi n..r. or
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pertinent systen paraaeters, discussed in Section II below, are within expected ranges.
The parameters of importance for these transients ir.clude reactor coolant system pressure, steam generator pressure, fluid temperatures, fuel and clad temperatures, discharge flow rate, steam line and feedwater flow rates, safety and relief valve flow rates, pressurizer and steam generator water levels, mass and energy transfer within the containment (for breaks inside containnent), reactor power, total core reactivity, hot and average channel heat flux, and minimum departure from nucleate boiling ratio (DNBR).
The sequence of events described in the applicant's safety analysis report (SAR) is reviewed by both RSB and EICSB. The RSB reviewer concentrates on the need for the reactor protecticn system, the engineered safety systems, and operator action to secure and maintain the reactor ir, a safe condi tion. The EICSB reviewer concentrates on the instrumentation and control aspects of the sequence described in the SAR to evaluate whether the reactor and plant pro-tection and safeguards controls cnd instrumentation systems wi.1 function as assumed in the safety analysis with regard to autonatic actuation, renote sensing, indication, control, and interlocks with auxiliary or shar systems. EICSB also _valuates potential bypass modes and the possibility of manual control by the operator.
The analytical methods are reviewed by RSB to ascertain whether the mathematical modeling dnd Corpute, codes have been previously reviewed and accepted by the staff. If a referenced inalytical method has not been previously reviewed, the reviewer requests initiation of a generic evaluation of the new analytical model by CPB.
In addition, the values of all the parameters used in the new analytical model, including the initial conditions of the core and systen, are reviewed.
APCSB reviews the auxiliary feedwater systen to verify that it can function following a feedwater line break, given a single active component failure and with either onsite or offsite power. This review is perfor ed as described in Stanoord Review Plan (SRP) 10.4.9.
RSS rcviews the auxiliary feedwater systen to confirn that the flow provided is acceptable for controlling the transient following a feedwater line break.
MEB evaluates potential water-hamrer effects on safety valve integrity.
AAS evaluates the fission product release assumptions used in deternining anv offsite releases.
AAB verifies that the radiological consequences resulting from a feedwater pipe creak are within acceptable limits. This analysis ;s done independently of the REB review.
CSB, under SRP 6.2.1, evaluates the response of the containment to bredks of feedwater lines with regard to the effects of pressure and terporsture on the containment functional capabilities.
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,00 15.2.8-2 if/
II.
ACCEPTANCE CRITERIA 1.
The basic objective of the review of feedwater system pipe break events is to confirm that the reactor primary system is maintained in a safe status for a range of feedwater line breaks up to and including a break equivaler.t in area to the double-ended rupture of the largest feedwater line.
2.
The specific criteria used in evaluating the consequences of these breaks are:
a.
Pressure in the reactor coolant and main steam systems should be maintained below 110; of the design pressures (Ref. 3).
b.
The potential for core damage should be evaluated on the basis that it is accept-able if the minimum DNBR remains above 1.30 or 1.32, as appropriate, based on cor-relations given in References 4 and 5.
If the DNBR falls below these values, fuel damage (cod perforation) should be assuned unless it can be shown, based on an acceptable fuel damage model, that no fuel failure results. If fuel damage is calculated to occur, it should be of sufficiently limited extent so that the cnre will remain in place and geometrically intact with no loss of core cooling capability.
c.
Any activity release must be such that the calculated doses at the site boundary are well within the guidelines of 10 CFR Part 100.
3.
There are certain assumptions which should be used in the analvsi', regarding important parreters that describe initial plant conditions and postulated system failures. These are listed below.
a.
The power level assumed and number of loops operating at the initiation of the transient should correspond to the operating condition which maximizes the conse-quences of the accident. These assumed initial conditions will vary with the par-ticular nuclear steam supply systen and sensitivity studies will be required to deternine the most conservative combination of power level and plant operating mode. These sensitivity studies may be presented in a generic report and refer-enced if considered applicable.
b.
The assumptions as to whether offsite power is lost and the tine of loss should be made conservatively. Offsite power may be lost simultaneously with the occurrence of the pipe break, the loss may occur during the accident, or offsite power may not be lost. A study should be made to determine the most conservative assumption appropriate to the plant design being reviewed. The study should take account of He effects that loss of offsite power has on reactor coolar,t and main feedwater pump trips and on the initiation of auxiliary feedwater, and the resulting modifi-cation of the sequence of events.
c.
The effects of the postulated feedwater line breaks on other systems (pipe whip, jet impingement, reaction forces, tenperature, humidity, etc.) should be considered in a manner consistent with the intent of Branch Technical Position APCSB.
and MEB 3-1 (Ref. 9).
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9
)}-1 15.2.8-3
d.
The worst single active tcmponent failure should be assumed to have occurred in the systems required to control the transient.
e.
The maximum rod worth should be assumed to be held in the fully withdrawn position.
An appropriate rod reactivity worth versus rod position curve should be assumed.
f.
The core burnup (time in cc e life) should be selec*ed to ' eld the most limiting combination of noderator temperatu e coefficient, void coef ficient, Doppler coef-ficient, axial power profile, and radial power dist ribution.
g.
The initial core flow assumed for the analysis of the feedwater line rupture acci-dent d.ould be chosen co"servatively. If the min mum core flow al wwed by the i
technical specifications is assured, the mininum % BR rargin results for the case of a feedwater line rupture inside containment. However, this may not be the most conservative assumption. For example, maximum initial core flow results in increased reactor coolant system cooldcwn and depressurization, decreased shutdown margi", and an increased possibility that the core will become critical and return to power. Since it is not c! car what initial core flow is most conservative, the applicant's assumotion should be justified by appropriate sensitivity studies.
III.
REVIEW PRCCEDURES The procedures below are used during reviews of both construction permit (CP) and operating license (CL) applications. During the CP review the values of systc-m para eters and set-points used in the analysis will be preliminary in nature and subject to change. At the UL review stage, final values shculd be used in the analysis, and the reviewer should corpare these to the limiting sa.ty system settings included in the proposed technical specifications.
The values of system paraTeters and initial core and system conditions used as input to the nodel are reviewed by RSB and are compared to the initial ccnditions listed in Section II of this plan.
Of particular importance are the reacti sity coef ficients and ccntrol rod worths used in the applicant's analysir and the variation of roderator temperature, void, and Doppler coef ficients of reactivity with core life. The justification provided by the applicant to shew that he has selected the core burnup that yieids the minimum margins is evaluatcJ. CPB is consulted renarding the values of reactivity parameters used in the applicant's analysis Analytical models should be of sufficient detail to simulate the reactor cool.at (prinary),
stean generator (secondary), and auxiliary systers. The equations, sensitivity studies, and rodels proposed by the apolicant are reviewed bj RSB to determine if these have been previously reviewed ard found acceptable by the staff. For situations where new generic rethods are proposed, the reviewer will request an evaluation by CFB. Acceptable equations, sensitivity studies, and rodels are described in References 6, 7, and 6.
mm
Credit taken for a reactor trip signal or for actuation of engineered safety features should be reviewed by EICSB to detemine the ability of the instrumentation and control systems to respond as assumed under accident conditions.
The ability of the auxiliary feedwater system to supply adequate feedwater flow to the unaffected steam generators during the accident and subsequent shutdown is evaluated by APCSB as to availability and by RSS as to capability to effect an orderly shutdown. Since auxiliary feedwater system designs are diverse and may require both automatic and manual actuation, preoperational tests should be specified to identify any necessary operator actions and to determine the maximum times permitted for their completion.
To the extent considered necessary, the RSB reviewer evaluates the effect of si " active failures of systems and components that may alter the cot.rse of the accident. This phase of the review uses the system review procedures described in the standard review plans for Chapters 5, 6, 7, 8, and 10 of the SAR. The variations with time during the transient of parameters listed in Secticas 15.X.X.3(C) and 15.k.X.4(C) of the Standard format (Ref.
- 2) are reviewed. The more inportant of these parameters for the feedwater line break accident (as listed in Section I of this SRP) are compared to those predicted for other similar plants to see that they are within the expected range.
The reviewer confirms that the amount of secondary coolant expelled from the system has been calculated conservatively by evaluating the applicant's methods and assumptions, by comparison with an acceptable analysis performed on another plant of similar design, or by comparison with staff calculations for typical plants which will be available from CPB on request.
The reviewer confirms that a comitment has been made in the SAR to conduct preoperational tests to verify that valve discharge rates and response times including, for example, opening and closing times (delay times) for main feedwater, auxiliary feedwater, turbine and main steam isolation valves, and steam generator and pressurizer relief and safety valves, have been conservatively modeled in the accident analyses. In addition, preoperational testing should include verification of reactor trip delay times, startup delay times for actuation of the auxiliary feedwater system, safety injection signal delay time, and delay times for delivery of any high concentration boron injection required to bring the piant to a safe shutdown condition.
Using the information developed in the review, the AAB reviewer evaluates the radiological consequences of the design basis feedwater line break. This evaluation based on a qualita-tive comparison with the results of the design basis steam line break, or on a detailed analysis using the approach described in the appendix to Standard Review Plan 15.1.5.
IV.
EVALUATION FINDINGS The reviewer verifies that the SAR contains suf ficient infomation anJ his review supports the following kinds of statenents and conclutions, which should be included in the staff's e
safety evaluatinn report:
15.2.8-5
"The analyses and effects of a spectrum of feedwater line breaks inside and outside ccqtainnent, during various modes of operation, and with or withaut of fsite powcr have been reviewed. The accident which resulted in the most severe consequences was deter-mined and evaluated using a mathematical model that had been previously reviewed and found acceptable by the staff. The values of the parameters used as innut to this model were reviewed and found to be suitably conservative. The results of the analyses of the spectrum of feedwater line break accidents showed that fuel damage was minimal and that no loss of core cooling capability would result. The minimum departure from nucleate boiling ratio experienced by any fuel rod was
, resulting in of the rods experiencing cladding perforation. The maximum pressure within the re3ctor coolant and main steam systems did not exceed 1103 of the design pressures.
"The radiological consequences of the design basis feedwater line break have been evaluated. Technical specification limits on primary and secondary coolant activities limit potential doses to small fractions of 10 CFR Part 100 exposure guidelines.
" Based on the above, the staff concludes that the plant design is acceptable aith regard to feedwater line break accidents.'
V.
REFERU.CES 1.
10 CFR Part 100, ' Reactor Site Criteria."
2.
Regulator, Guide 1.70, " Standard Fornat and Content of Safety Analysis Re? orts for Nuclear Power Plants, Revision 2.
3.
ASME Boiler and Pressure Vessel Code,Section III, " Nuclear Power Plant Components,'
Article NB-7000, " Protection Against Overpressure,' A:nerican Society of Pechanical Engineers.
4 L. S. Tono, " Prediction of Departure trom Nucleate Boiling for an Axially Non-Uniform Heat Flux Distribution, Journal of Nuclear Energy, Vol. 21, 241-243 (1967).
5.
" Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water,' Babcock &
,4ilcox Power Generation Systems, BAW-10000, Supplerent 1, March 1971.
6.
' Reference Safety Analysis Pepo t RESAR-3,' West aghouse Nuclear Energy Systems, Novec ber 1973; and " Reference Safety Analysis Repor* - RESAR-41," Westinghouse Nuclear Energy Systems. recerter 1973 (under review).
7.
" System 80 Standard Safety Analysis Report (CESSAR), Combus' on Engineering, Inc.,
"ugust 1973 (under review).
8.
" Standard Nuclear Steam Systen B-SAR-241,' Babcock & Wilcox Company, february 1974 (under review).
15.2.8-6
9.
Branch Technical Positions APCSB 3-1, " Protection Against Postulated Piping Failures in Fluid Systems Outside Containrent," attached to SRP 3.6.1 and MEB 3-1, " Postulated Break and Leakage Locations in Fluid System Piping Outside Centainment," attached to SRP 3.6.2.
6 15.2.8-7