ML19221B140
| ML19221B140 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-15.1.5, NUREG-75-87, NUREG-75-87 15.1.5, SRP-15.01.05, SRP-SRP-15.01.05, NUDOCS 7907120503 | |
| Download: ML19221B140 (12) | |
Text
N U R E G-75/087
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M[Vih STANDARD REVIEW PLAN M
UrS. NUCLEAR REGULATORY COMMISSION
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OFFICE OF NUCLEAR REACTOR REGULATION SECTIO 1 i5.1.5 SPECTRUM OF STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT (PWR)
REVIEW RESPONSIBILITIES Primary - Reactor Systens Branch (RSB)
Secondary - Accident Analysis Branch ( AAB)
Auxiliary and Power Conversion Systems Branch (APCSB)
Containa nt 5ystems Cranch (CSB)
Core Perfor~ance Branch (CPB)
Electrical, Instrumentation and Control Systems Eranch (EICSB)
Mectanical Engineerir+g Branch (MES) 1.
AREAS OF REVIEW The steam release resulting from a rupture of a main stean pipe will cause an increase in steam flow which decreases with time as the steam pressure decreases. The increased steam flow causes increased energy renoval fron the reactor coolant systen and results in a reduction of coolant temperature and pressure. Due to the negative moderator temperature coefficient this coo'iown causes an increase in core reactivity. The core reactivity increase causes a pc wer level increase and a decrease in shutdown narg!n. If the plant is at power the reactor is automatically tripped nd the main steam and feedwater line isola-e tien valves are automatically closed. Decay heat is removed through the unaffected ste3n generators by venting steam from the secondary system safety and relief valves. The auxiliary feedwater system supplies makeup water to the unaffected stean generators.
The transient following a stcan line break is sensitive to the discharge rate so that a range of break si n s must be evaluated both inside and outside containment to deternine the accept-ability of the system response. The course the transient takes and its ultimate effects also depend on the assumed initial power level and mode of operati)n (i.e., hot shutdown, full power, one, two, or three-loop operation). Analyses with various assumed initial conditions are required to verify that the condition leading to the var 4s; c r) equences has been tl/ l l M)
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I V
The topics reviewed include: postulated initial core and reactor cor.ditions pertinent to the stean line break accidente nethods of thermal and hydraulic analyses, postulated sequence of events including analyses to determine the time c' reactor trip and time delays prior to and subsequent to initiation of the reactcr protection systen, assumed responses of the reactor coolant and auxiliary systens, functional and operational characteristics of the USNRC STANDARD REVIEW PLAN Stendard review piene ore propered for the guidence of the Of6ce of Nue' ear Reactor Reguleoon staff responsible f or the row,ew of apphcetions to construct end opere o aucteer power piante These documente are made available to the pubhc es part of the Commies.on i pohey to mform the nucteer industry and the generes pubhc of regutetory procedus es and pohcies Standard review piene are not oubetetutes for reguletory guideo or the Commise+on e regulatione end comphonce wa-h them se not required The evendeed review pien sectione. ere keved to Revision 2 of the Stenderd Forenst end Conten* of Sofety Analve o Reper's for N.scieer Powee Plante Not en sections of the Stenderd Formet have e corresponding review plan Pubbebed o'endard review plane writ be revised pereedsco y se appropeieie to accc wmodate commente end to refioct new informet+on and esperiente n
Commeate and suggestione for improvement wui be considered and should be sent to the U S Nucieer Reguie'ory Commise.on Office of Nuc!*er Reactor Regwletion Weehengton. D C 4265 iI f (/
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reactor protection systen in terns of its ef fects on the sequence of events, and operator actions required to secure and maintain the reactcr ir a safe shutdcwn condition.
The results of the analyses are reviewed to ensure thdt pertinent systen paraneters are within expected ranges. The parameters of importance for these transients include reactor coolant system (RCS) pressurt, stean generitor pressure, fluid temperatures, fuel and clad temperatures, discharge flow rate, stean line and feedoater flow rates, safety and relief valve flow rates, prr,surizer and steam generator water levels, nass and energy transfer within the contai rent (for break s inside containment), reactor power, total core reactivity, hot and a terage channel heat flux, and nininun departure fron nucleate boiling ratio (DZR).
The sequence of events descrited in the applicant's safety ar.alysis report (SAR) is reviewed by both RSB and EIC5B. The RSB reviewer concentrates on the need for the reactor protec-tion systen, the engineered safety systems, and operator action to secure and maintain the reactor in a safe condition. The EICSB reviewer concentrates on the instrumentation and controls aspects of the sequence described in the Sf.R to evaluate whether the reactor and plant protection and safeguards controls and instrumentation systens will function as assumed in the safety analysis with regard to automatic ettu nion, renote sensing, indica-tion, control, and interlocks with auxiliary or shar2d systens. EICSB also evaluates poten:ial bypass nodes and the possibility of manual control by the operator.
The analytical rethods are reviewed by RSB to ascertain whether the mathematical modeling and computer codes have been previously reviewed and accepted by the staff. If a referenced analytical nethod has not been previously reviewed, the reviewer requests initiation of a generic evaluation of the new analytical nodel 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-AFCSB reviews tne auxiliary feedwater systen to see that it can function following a stean line break given a single active conponent f 5 e with eit W cosite or offsite power.
This is done as described in Standard Review Plan (SRP) 10.4.9.
RSB reviews the auxiliary feedwater systen to see that the flow provided is acceptable for controlling the transient following a stean line break.
MEB evaluates potential water-hanner effects on safety valve integrity.
AAB evaluates the fiss;cn oroduct release and verifies that the radiological consequences resulting fran a stean line break are within acceptable limits This evaluation is per-fomed for the design basis case as described in the appendix to this review plan.
CSB, as described in SRP 6.2.1, evaluates the response of the containnent to ruptures of steam lines with regard to the effects of pressure and temperature on the containment functicnal capabilities.
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II.
ACCEPTANCE CRITERIA 1.
The general objectis? Of the review of stean lire rupture events is to confirn that the primary reactor coolant systen is naintained in i safe status for a range of steam line ruptures up to and including a break equivalent in area to the double-ended rupture of the largest steam line.
2.
The specific criteria against which the consequences of these breaks are to be evaluated are:
a.
Pressure in the reactoc coolant and main stean systems should be maintained below 110 of the design pressures (Ref. 2).
b.
The potential for cor(.amage is evaluated on the basis that it is acceptable if the minimun DNBR remains above 1.30 or 1.32, as appropriate, based on correlations given in References 3 and 4.
If +he CNBR falls below these values, fuel damage (rod perforation) must be assumed unless it can te shown, based on an M ceptable fuel da age model, that fuel failure has not occurred. Any fuel damage calculated to occur nust be of sufficiently linited extent that the core will remain in Llace and intact with no loss of core cooling capability.
c.
The radiological criteria used in the evaluation of stean systen pipe break accidents ( MR's only) appear in the appendix to this plan.
3.
Tnere are certain assunctions regarding important parameters used to describe the initial plant conditions and postulated systen f ailures which should be used. These are listed below:
a.
The reactor power level and number of operating loops assumed at the initiation of the transient should correspond to the operating condition which naxinizes the consequences of the accident. These assumed initial conditions will vary with the particular nuclear stean supply systen (1555) design, and sensitivity studies will be required to deternine the most 'onservative combination of power level and plant operating rode. These sensitivity studies may be presented in a generic report and referenced in the SAR.
b.
Assumptions as to the loss of of fsite power and the tire of loss should be nade so as to naxinize the consequences of the accident. A loss of offsite poser nay occur sinultaneously with the pipe break, or during the accident, or offsite power nay not be lost. Analyses should be made to determine tha nost conservative assu ption appropriate to the particular plant design. The analyses should take account of the effect that loss of offsite power has on reactor coolant pu~p and main feed-water pump trips and on the initiaticn of auxiliary feedwater flow, and the effects or, the sequence of events for these accidents.
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c.
The effects (pipe whip, jet inpingerent, reaction forces, temperature, humidity, etc.) of postulated steam line breaks on other systens should be considered in a manner consistent with the intent of Branch Technical Positions APCSB 3-1 and MEB 3-1 (Ref. 8).
d.
The worst single active component failure st.ould be assumed to occur. The assumed single failure may cause nore than one stean generator to blow down, or may oe in any of the systens required to control the transient.
e.
The naximum-worth rod should be assumed to be held in the fully withdrawn position.
An appropriate rod reactivity worth versus rod cosition curve should be used.
f.
The core burnup (time in core life) should be selected to yield tte most liniting combination of noderator tenperature coefficient, void coef ficient, Doppler coefficient, axial power profile, and radial power distribution, g.
The initial core flow assumed for the analysis of the steam line rupture accident should be chosen conservatively. If the minimun core flow allowed by the techni-cal specifications is assumed, the ninimum DNBR nargin results, but for the analysis of steam line break accidents this nay not be the most conservative assu'? tion. For example, maxinum initial core flow results in increased reactor coolant system cooldown and dcpressurization, decreased shutdown margin, and an increased possibility that the core will becore critical and return to power.
Since it is not clear what initial core flow is most conservative, the assumed value should t'e justified.
III.
REVIEW PROCEDURES Tne procedures below are uced during both the construction permit (CP) and operating license (DL) reviews. During the CP review, the values of systen parc~.... and setpoints used in the a nalysis will be preliminary in nature and subject to change. At the OL review stage, final values should be used in the analysis, and the reviewer should compare these to the limiting safety syttem settings included ir '.he proposed technical specifications.
- 1. The reviewer determines the acceptability of the analytical models and assumptions as follcws:
a.
The values of system parameters and initial core and system conditions used as input to any analytical model are reviewed by RSB. Of particular importance are the reactivity coefficients and control rod worths used in the analysis, and the variation of c'oderator temperature, void, and Doppler coefficients of reactivity with core life. The justification provided by the applicant to show that he has selected the core burnup that yields the minimum margins is evaluated. CPB is consulted regarding the values of the reactivity parameters used in the analysis.
The reviewer confirms that the amount of secondary coolant expelled from the system (for breaks outside containment) has been c._lcuiated conservatively by evaluating 15.1.5-4
the applicant's methods and assumptions, by comparing with an acceptable analysis performed on another plant of similar design, or by comparing with staff calcula-tions for typical plants done by CPB on request.
b.
The acceptability of the equations, sensitivity studies, and models proposed by the applicant are evaluated. For situations where new generic Fethods are proposed, the reviewer will reauest an evaluation by CPB. Acceptable equations, sensitivity stu1ies, and nodels are dese.ribed in References 5, 6, and 7.
c.
Analytical models should be sufficiently detailed to simulate the reactor coolant (primary), steam generator (secondary), and auxiliary systems. The reviewer evaluates the following functional requirements:
(1) Rcactor trip signal: credit takea for any reamtor trip signal is reviewed by ETCSB to confirn that, under accident conditions, the instrumentation and control systems are capable of the aseumed response.
(2) Emergency core cooling system (ECCS): credit taken for actuation of the ECCS is reviewed by EICSB to verify the ability of the instrumentation and control systems to respond as assu~ed.
(3) Auxiliary feedwater system: the availability of the auxiliary feedwater system to supply adequate auxiliary feedwater flow to the intact stean generators during the accident and the subsequent shltdown condition is evaluated. This is done by APCSB as to availability of the system and by RSB as to capability to effect an orderly shutdown. Since auxiliary feedwater system designs are diverse and may require both autonatic and manual actu3 tion, preoperational tests should be specified to iJentify any necessary operator actions and to establish times required for their completion.
d.
The variations with time during the transient of parameters listed in Sections 15.X.X.3(c) and 15.X.X.4(c) of the Standard Format (Ref. 1) are reviewed. The values of the more important of these parameters for the steam line break accident (as listed in Section I of this SRP) are compared with those predicted for other similar plants to see that they are within the range expected.
2.
To the extent deered necessary, the reviewer evaluates the effect of single active failures of systems and co~penents that ray affect the course of the accident. Inis phase of the review is done using the systen review procedures described in the SRP for Chapters 5 6, 7, 8, and 10 of the SAR. The reviewer also considers single failures that nay cause more than one steam generator to blow down, thus increasing the reactiv-ity addition to the core.
3.
The reviewer Confirms that a comnitment has been made in the SAR to conduct preoperational tests for verifying that valve di3 charge rates and response tir.es (including, for example, opening and closing times for main feedwater, auxiliary feedwater, turbine and main steam
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15.1.5-5
isolation valves, and steam generator and pressuri7er relief and safety valves) have been conservatively nodeled in the accident analyses. In addition, preoperational testing should include verification of reactor trip delay tines, startup celay tiaes for auxiliary feedwater system actuation, safety injection signal delay tire, and delay tires for delivery of any high concentration boron solution requir ed to bring the plant to a safe shutdo.sn condition.
4 Based on the above infornation, AAB evaluates the radiological consequences of the design basis steam line break accident as described in the appendix to this plan.
IV.
EVALUATION FINDINGS The reviewer verifies that the SAR contains sufficient information and his review supports the following kinds of statements and conclusions, which should be included in the staff's safety evaluation report:
"The analyses and effects of steam line break accidents inside and outside containment, oming various modes of operation and with or without of f-ite power, have been reviawed.
The accident which resulted in the nost severe consequences was deternined and evaluated using a mathematical nodel that had been previously reviewed and found acceptable by t% staff. The para eters used as input to this model were reviewed and found to be su itably conservative. The results of the analys:s of the spectrum of steam line break accidents showed that fuel damage was ninimal and that no loss of core cooling capability resulted. The minimum departure from nucleete boiling ratio (DNBR) experienced by any fuel rod was
, resulting in of the rods experiencing clad perforation. The maximun pressure within the re3ctor coolant and main steam systems did not exceed 110' of the design pressures.
"The radioactivity release has been evalua'ed usirg the computer code LARA and a con-servative description of the plant response to the accident.
A decontamination factor of betwtan the water ard steam phasec and a 3
X/1 value of sec/n has been used in our evaluation of radiological consequences.
The calculated doses are presented in Table
. Technical specification limits on primary and secondary coolant act.ivities will limit potential doses to small fractior.s of the IJ CFR Part 100 exposure guidelines. The potential doses are within the 10 CFR Part 100 exposure guidelines even if the accident should occur coincident with an iodine spike."
" Based on the above, the staff concludes that the plant design is acceptable with regard to steam line break accidents."
V.
REFERENCES 1.
Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," Revision 2.
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2.
ASME Boiler and Pressure Vessel Code,Section III, " Nuclear Power Plant Conponents,"
Article NR 7000, "Piotection Against Overpressure," American Society of Mechanical Engineers.
3.
L. S. iong, " Prediction of Departure from Nucleate Boiling for an Axially Non-Uniform Heat Flux Distribution " Journal of Nuclear Energy, Vol. 21, 241-248 (1967).
4.
" Correlation or Critical Heat Flux in a Bundle Cooled by Pressurized Water," Babcock
& Wilcox Pover Generation Systems, 3AW-10000, Supplement 1, thrch 1971.
5.
" Reference Safety Analysis Report - RESAR-3," Westinghouse Nuclear Energy Systems, November 1973; and " Reference Safety Analysis Report - RESAR-41," Westinghouse Nuclear Energy Systens, December 1973 (under review).
6.
" System 80 Standard Safety Analysis Report (CESSAR),' Ccmbustion Engineering, Inc.,
August 1973 (under review).
7.
" Standard Nuclear Steam Supply Syste, B-SAR-241," Babcock and Wilcox Company, February 1974 (under review).
8.
Brarch Technical Positions APCSS 3-1, " Protection Acainst Postulated Piping Failures in Fluid Systems Outside Containrient," attached to SRP 3.6.1, and MEB 3-1, " Postulated Break and Leakage Locations in Fluid Sys ten Piping Outside Containr'ent," attached to SRP 3.6.2.
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APPENDIX STANDARD REVIEW PL AN SECTION 15.1.5 RADIOLOGICAL CONSEQUENCES OF MAIN STEAM LINE FAILURES OUTSIDE CONTAINMENT (PWR)
RE V I E W R E_S PtJNS I_B ILj T I E_5 Prir:ary - Accident Analysis Branch (AAB)
Secondary - Hydrology and Meterology Branch (HMD)
Reactor Systens Branch (RSB)
Ef fluent Treatnent Systems Branch (ETSB)
I.
AREAS OF REVIEW The AAB reviews (a) the sequence of events in the applicant's description of the steam line failure accident outside containment, with and without of fsite power, to assure that this sequence of events represents the nost severe case from the standpoint of release of radioattive raterials, and (b) the coolant activity concentration technical specification linits to assure that potential doses resulting from this accident are adequately limited.
The HMB provides acceptable atmospheric dispersion X/Q values for this accident. The ETSB provides acceptable rudels and assumptior3 for iodine spikir.g and their ef fects upun coolant activity. The RSS determines the acceptability of the applicant's descriptien of events, including operator actions, for this accident.
II.
ACCEPTANCE CPITERIA Standard Technical Specification (STS) limits on PWR prirury and secondary coolant activity concentrations and primary-to-secondary leak rate have been issued (Ref erences 4, 5 and 6).
The plar.t is censidered adequately designed against the consequences of a rain steam line failure outside containment if calculations show that the resulting doses at the exclusion area and low population zone boundaries, based on ST5 limits, are: (a) small fractions (less than 101) of the la CFR Part 100 exposure guidelines, and (b) within 10 CFR Part 100 gaidelines for 'che cases of a preaccident iodine spike or one rod held out of the core.
If the doses are not within these guidelines using the STS limits, the technical specific -
tion limits on coolant cone.entrations and/or primary-to-secondary leak rate are reduced urtil the calculated doses are within these guidelines.
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III. REVIEW PROCEDURES The reviewer selects and erphasizes aspects of the areas covered by this appendix as may be appropriate for a particular case. The judg ent on areas to be given attention and emphasis in the review is based on an inspection of the raterial presented to see whether it is similar to that recently reviewed on other plants and " ether items of special saf ty significance are involved.
At the construction permi' stage, there is generally insuf ficient ir fonration available tt make meaningful radiological conseqJence calculations for -hese accidents. At this stage, the review is limited to a brief review of the applicant's discussion of the main steam line f ailure accidents to deternine that there are no unusual design f eatures that would preclude the limitation of radiological consequen;es by appropriate linits on coolant con-centrations and primary-to-secondary leakage. The detailed review of radiologica! con-sequences of the rain steam line f ailure accidents is done at tSe operating license stage when system parameters are fully developed.
The Standard 7 chnical Specifications for each of the three PWR vendors' NSSS include limits on the primary and secondary coolant activities and primary-to-secondary leak rate.
If the applicant proposes to use these standard limits and his plant is one of the standard NSSS/ BOP plants fcr which the steam line failute accident has been evaluated generically with the standard coolant activity and leakage limits, the reviewer need not reevaluate tt; o.fsite doses from this accicont if the X/Q's for the site under review are lower than the lii;iting X/Q used in the generic review of the standard plant steam line failure.
The AAC revjew of rain steam line failure accidents at the operating license st-consists of the following:
1.
Review of the applicant's descriptions of the steam line failure accident (with and wi thout of fs i te power). This includes a review of the time scquence of occurrence of events.
2.
The RSB should be contacted for a determination of the acceptability of the applicant's description of events including operator actions. The AAB reviews the sequence of events to asscre that this sequence of events represents the most severe case from the standpoint of release of radioactive raterials and calculated doses.
3.
Determination of coolint activity contentrations. The reviewer assures the primary and secondary coolant activity concentrations allowed by the technical specifications (SAR Chapter 16 or tre Standard Technical Specifications given in References 4, 5 or 6) as equi'ibrium conditions prior to the accident.
4 Deterr.ination of the iodine spiking effects. Two cases are analyzed, one with an 9
iodine spike assured to begin because of re3ctor trip or primary system depressuriza-tion when the steam line break occurs, and one with an iodine spike assumed to have bequn well before the accident due to a previous reactor transient (Refs. 3 and 7).
9 Rcv. I 15.1.5-9
The ETSB provides the AAB with the iodine spiking rodel.
If the applicant proposes to use a spiking rodel different than the rodel descrited below, the ETSC 2hould te requested to review the model.
The current. spiking rodel assures that at the tir.e the stear lire break occurs, the iodine release rate from the fuel rods to the primary coolant (expressed in unit of curies per unit time) increases by a factor of SCO relative to the release rate cal-culated assuming that the prirary coolant iodine concentration is at the equilibriuc concentration technical specification limit.
For the case with an iodine spike which already exists (due to a previous transient),
the iodine concentration is assured to be that allov.ed oy Figure 3.4-1 of References 4, 5 or 6, or that which is preposed in Chapter 16 of the SAR.
5.
Review of the ef fects of possible fuel damage during the accident on exclusion area boundary and LFZ doses. Additional coolant activity ray becorre availat,la for release if fuel failures result frcm the accident. The PSS reviews the effect of a s tean line break (with the rost reactive control rod stuck at its fully withdrawn pesition) on cure thermal argins. If this event is predicted to cause fuel failures, RS6 notifies R
The reviewer assu es the applicant's calculations of fuel de~ age are correct unless inforced otheruise by RSS.
If fuel daTage does occur, calculations should be perforred in order to assure that 10 CFR Part 100 guidelines are not exceeded (with-out a preaccident iodine spike).
6.
Determination of the leakage into the steam generators. Norra l operating prima ry-to-secondary leakage is assured to exist in the steam generators. The leakage rate should be the raximum allowed by the techn cal specifications. This value is 1 gpm i
in the STS but n'ay be icwer if required bt se of the ccnsequences of a rod ejaction xcident or an anticipated transient withei.
scram (ATWS). The leakage should te apportioned between affected and unaffected steam generator (s) in such a manner that the calculated dose is maximized.
7.
Cetermination of iodine transport. During pericis of stea generator dry-out, all iodine transported to the secor.dary side by prirary coolant leakage is assumed to be rtleastd to the environment. Curing periods of total submercence of the tut:es, the fraction of iodine lost is equal to the flash fraction of the primary coolant leakage; appropriate credit for scrubbing by the secondary coolant nay also be claired (Peference 2).
Any iodine transferred to the secondary side coolanc will tecome air-borne at a rate which is a function of the steaming rate end iodine partition coefficient. An iodina partition coefficient of 100 between steam generator water and steam phases may be conservatively assumed, unlest the applicant presents reasonable evidence that the use of sone other value is jus +ified.
8.
The AAB provides the HY3 with the release location and release conditions. The HMS ther provides the AAB with appropriate X/Q values for this accident.
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9.
Ca'culation of the exclusion area boundary (EAB) and low population zone (LPZ) boundary doses. The reviewer computes the doses for the steam line break accident, both with and without preaccident iodine spiking. A breathing rate of 3.47 x 10 rd/sec is used in the calct.lation of thyroid doses for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the steam line break.
10.
Review of the results of the dose calculations. The calculated doses are considered acceptably low if: (1) the doses calculated without assuming the existence of a,
accident iodine spike, but assuming that iodine spiking occurs as a result of the accident, are less than a small f raction of 10 CFR Part 100 guidelines; and (2) the doses calculated assuming preaccident iodine spiking or 3dditional fuel failures occur as a result of the accident (assuming the most reactive control rod remains stuck in its fully withdrawn position) are less than the dose guidelires of 13 CFR Part 100.
If the doses are not within these gadelines, the technical specification limits on ec4uilibrium and,or spiked primary coolant activity, or the teChLiCal specification limit on primary-to-secondary leak rate, should be reduced until the Calculated doses are within these guidelines.
f IV.
Ei All'ATION FINDINGS The reviewer verifies that suf ficient inf ormation has been proviced and the review and calculations support conclusiens of the following type, to be included with the R5B find-ings in the staff's evaluati n report at the operating license stage:
"The radioactivity release has teen evaluated according to a conservative descr!ption of the plant response to the accident. Our calculated doses are presented in Table and our assumptions are listed in Table
'iechnical specification limits on prinary and secondary coolant activities will limit potential doses tc small fractions of the 10 CFR Part 100 exposure guidelines. The potential doses are within the 10 CFR Part 100 exposure guidelines even if the acci-dent should occur with a preaccident iodine spike or assuming additional fuel failures occur during the accident as a result of the most reactive centrol rod reraining fully withdrawn."
The following paragraph is added if fuel damage is found to be a possible consequence of the accident:
'ine evaluation of the main steam line failure outside ccntainr.ent has been evaluated with fuel drage in the core (as a result of the most reactive control rod remaining fully withdrawn). The resulting doses are within the guidelines of 10 CFR Part 100 orovided the normal operating primary-to-secondary leakage is limited to got "
At the construction permit stage, the following paragraph is included with the RSB findings 9
in the staft's safety evaluaticn report:
Rev. I 15.1.5-11
"On the basis of our experience with the evaluation af steam line and steam generator tube failure accidents for PWR plants of similar sesigr., we have concluded that the consequences of these accidents can be control'ed by limiting the permissible primary and secondary coolant system radioactivity cancentrations 3nd/or primary to secondary leak rates so that potential offsite doses are small. At the operating license stage, we will include appropriate limits on these parameters to be included in the plant technical specifications."
If the plant is a standard plant which has been reviewed before, the following paragraph may be used:
"The radiological consequences of a steam line break in the (vendor's naye) standard USSS/(A/E's name) BOP were analyzed generically by the staff and the results reported in the staff SER dated (NSSS/ SOP SER issue date). The offsite doses were found to be acceptable for sites with X/Q's equal to or less than sec/m, if the voolant activities and steam generator primary-a-secondary leak i' ate are limited to the values in the (vendor's name) Standara Technical Specifications (NUREG-(issue date)). The technical specifications the applicant will use include these coolant activity and steam generator leak limits. The staff has estimated the 0-2 hour X/Q at the exclusion area boundary of the site is sec/m, as discussed in Socticn (2.3.4) in this SER.
This X/Q is lower than the limiting X/Q for this standard NSSS/ BOP plant. Therefore, we conclude that the offsite doses from a steam lir.o break accident in the plant would be less than the current dose guidelines and are acceptable, although a specific dose calculation for this accident has not been perforned."
V.
REFERENCES 1.
10 CFR Part 100, "Peactor Sitc Criteria."
2.
A. K. Postra anc ~
S. Tam, " Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident," huREG-0409, USNRC, 1978.
3.
R. R. Bellamy, "A Peaulatory Viewpoint of Iodine Spiking During Peactor Transients,'
Trans. Am. Nucl. Soc., 28 (1978).
4.
Standard Technical Specifications for Cortustion Engineering PWRs, NLREG-0212.
5.
Standard Technical Specifications for Westinghouse PWRs, hUREG-0452.
6.
Standard Technical Specifications for Babcock and Wilcox PWRs, NUPEG-0103.
7.
W. F. Pasedag, " Iodine Spiking in GWR and PWR Coolant Systems," C0lf-770708, 3-217 (1977).
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15.1.5 lE Rev. 1