ML19221B149
| ML19221B149 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| 15.03.02, NUREG-75-087, NUREG-75-087-15.3.1, NUREG-75-87, NUREG-75-87 15.3.1, SRP-15.03.01, SRP-SRP-15.03.01, NUDOCS 7907120529 | |
| Download: ML19221B149 (6) | |
Text
NU REG-75/087 pa af cg4 f
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[i v't U.S. NUCLEAR REGULATORY COMMISSION bhh)
STANDARD REVIEW PLAN
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OFFICE OF NUCLEAR REACTOR REGULATION SECTION 15.3.1 LOSS OF FORCED REACTOR COOLANT FLOW INCLUDING 15.3.2 TRI. OF PUMP AND FLOW CONTROLLER MALFUNCTIONS REVIEW RESPONSIBILITIES Frimary - Reactor Systems Branch (R58)
Secondary - Cor rformance Branch (CPR)
Electr. cal, Instruwntatiu, and Control Systems Branch (EICSB) 1.
ARFAS OF P' m W A decrease in reuctcr cus
. floc occ,ing while the plant is at power could result in a degradation of core heat transfer. The resulting increase in clad temperature could result in fuel damage. A number of transients that are expected to occur with moderate freauency and that result in a decrease in forced reactor cooiant flow rate are covered by this re-view plan. Each of these transients should be discussed in individual sections of the applicant's safety analysis report (SAR), as required by the Standard Format (Ref.1).
Core thermal and hydraulic transients associated with n rtial and corplete loss of reactor coolant flow are evaluated. Tiese include:
1.
For boiling water _ actors (EWR's), partial and complete recirculatior pump trips and malfunctions of the recirculation flow controller to cause decreasing flow.
2.
For pressurized water reactors (FWR's), partial and complete reactor coolant pump trips.
A partial loss of coolant flow nay be caused by a mechan: cal or electrical failure in a pump, a fault in the power supply to the pump, a pump trip caused by such ancialies as over-current or Mase imbalance, or a failure within the recirculatico flow control network (BWR) resulting in decreasing flcw. A corplete loss of fnrced coolant flow may result f rom the simultaneous loss of electrical power to all purps.
The review includes tne postulated initial core and reactor conditions which are pertinent to the loss of flow transient, the methods of therr31 and hydraulic analysis, the postulated sequence of events including time delays prior to and af ter protective system actuation, assured reactions of reactor systens components, tbc functional and operational character-istics of the reactor protection system ir terris of how it af fects the sequence of events, and all operator actions required to secure and maintain the reactor in a safe condition.
'7 3 0 71 ) b ( b USNRC STANDARD REVIEW PLAN st.r.de,d row ew piene er. p,epe,od fo, toe gu.dence of t*. owice of Nuci.e, R.scio, Reguistion sto e r ponsib e f, the row.w of appeicer+ons to construct and r
eponete nuccese power pleeste Theme documents are made evoetebte to the publee se part of the Comm soson e pokcy to soform the nuc6eer 6ndustry end the ge ieret pubHe of regutetory procedures and pohctee Stenderd rewtew plane are not subetetutwo for regulatory guedes or the Commasseon a regulatione end cornphonce w*th them as not requered The standeed review pian sections are keyed to Revieson 2 of the Standard Format and Content of Sofety Ano8vs.e Repo,ge 109 %ucteer Power Ptents Not ett sections of the Stenderd Formet have a corroepon ting review plan Pubhohed stenderd review piene will be rows,ed pe<+odicotiv. es oppropnete to accomm odete commente end to refioct new triformation and empenence Coenmente end suggestione for 6mpeowoment will be conside<ed and should be sent to the U $ Nucteer Regulatory Commessaan 0+fice of Nucieer Reactor Regutetson. Weshengton. O C 2%EE f
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The results of the applicant's flow transient analyses are reviewed to ensure that values of pertinent system parameters are within expected ranges for the type and class of reactor under review. These parameters include: peak clad temperature, peak fuel temperature, core flew and fl9w distribution, channel heat flux (average and hot), minimum critical heat flux ratio (or minimum critical powar ratio), departure from nucleate boiling ratio, vessel water level, themal power, vessel pressure, steam line pressure (BWR), steam line flow (BWR), and feed-water ficw (BWR).
The sequence of events described in the SAR is reviewed by both RSB and ElCSB. The RSB reviewer concentrates on the need for tt4 eactor protection system, the engineered safety system, and operator action to secure and maintain the reactor in a safe condition. The EICSB review, as described in Standard Review Plans (SRP) 7.2 and 7.3, concentrates on the instrumentation and controls aspects of the sequence described in the SAR to evaluate whather the reactor and plant protection and safeguards controls and instrumentation systems will function as assumed in the safety analysis with regard to automatic actuation, remote sensin1, indication, control, and interlocks with auxiliary or shared systens. EICSB also evaluates potential bypass modes and the possibility nf manual control by the operator.
The analytical rethods are reviewed by RSB to ascertaic whether the rathematical modeling and computer codes have been previousl-reviewed and accepted by the staff. If a referenced an-alytical method has not been prm, iously reviewed, the reviewer t equests initiation of a generic evaluation of the new analyt. cal model by CPB.
CPt3, as described in the appendix to SRP 4.4, perfonrs generic reviews of the thermal-hydraulic computer models used for this transient.
CPB also perforns, upon request, additional analyses related to these accidents for selected reactor types.
The values of all parameters used in a new analytical model, including the initial conditions of the core and system, are reviewed. It is the respoasibility of the RSS engineer to contact his counterpart in CPB to ensure that the relevant physics data have been used in any staff calculaticn; II.
ACCEPTA7CE CRITERIA 1.
The basic objectives of the review of loss of forced roactor coolant ficw transients are:
a.
To identify which of the transients are the most liniting.
b.
To verify that, fcr the most lit.iting trunsients, the plant responds to the loss of flow transients in such a way that the criteria regarding fuel damage and system pressure are ret.
2.
The specific criteria for incidents of moderate frequency
- are:
a.
Fressure in the Na; tor coulant and main steam systems should be raintainca below 110 of the ucsigr. pressures.
(Ref. 2).
7 The term "noderate frequency" is used in this review plan in the same sense as in the descriptions of design and plant process conditions in References 7 and 8.
15.3.1-2
b.
Fuel cladding integrity should be maintained by ensuring that acceptance criterion l of SRP 4.4 is satisfied throughout the transient.
c.
An incident of moderate frequency shculd not generate a more serious plant condi-tion withoJt other faults occurring indeperJently.
d.
An incident of roderate frequency in combination with any single active component failure, or single operator error, should not cause loss of function of any barrier other than the fuel claddir,]. A limited numbe-of fuel rod cladding perforations is acceptable.
3.
lhe applicant's analysis of the loss of reactor coolant flow tiansients should use an acceptable analytical model. The equations, sensitivity studies, and models descrit'ed in References 3 through 6 are acceptabla. If other analytical methods are pr oposed by the applicant, these methods are evaluated by the staff for acceptability. F)r new generic methods, the reviewer requests an evaluation by CPd.
The values of parameters used in the analytical mcdel should be suitably conservative.
The use of the following values is considered acceptable:
a.
The reactor is initially at rated output (licensed core thermal power) for the number of loops assumed operating, plus 25 to account for power measurerent uncertainty, b.
Conservative scram characteristics are assuned, i.e.
naximum time delay with the nost reactive rod held out of tFC core.
The core burnup is selected to f eld the most liniting conbination of moderator i
c.
temperature coefficient, void coefficient, Doppler coefficient, axial power profile, and radial power distritJtion.
II.
REVIEW PROCEDURES The procedures below are used during botn the construction permit (CP) and operating license (OL) reviews. During the CP review the values of system parameters and netpoints used in the analysis will be preliminarf in qature and subject to cnange. At the OL review stage, final values should be used in the analysis, and the reviewer should compare these to the limiting safet> system settings im iuded in
.e proposed technical specifications.
The description of each of the ioss of reactor coolant flow transients presented by the applicant in the SAR is reviewed by RSG regarding the occurrences leading to the initiating event. The sequence of events from initiation until a stabilized condition is reached is reviewed to ascertain:
1.
The extent to which norrally operating plant instrumentation and controls are assumd to func' ion.
'5.3.1-3 149 508
2.
The extent to which clant and reactor protection systers are required to function.
3.
The credit taken for the functioning of normally operating plant systems.
4.
The operation of engineered safety systens that are required.
5.
The extent to which operator actions are required.
If the SAP states th3t a particular loss of flow transient is not as limiting n sone other similar transient, the reviewer evaluates the justification presented by the applicant.
The reviewer confirms that all types of tiow loss transients are considered, e.g., purp trips du-ing two, three-and four-loop operation. The applicant is to present a quanti-tative ; ulysis in the SAR of the loss of flow transient that is detemined to be rost liniting. For this tr ars ient, the RSB reviewer, with the aid of the EICSB reviewer, reviews the timing of the initiation of those protection, engineered safety, and other systems needed to adequately limit the ccnsequerces of the loss of flow. The R5B reviewer corpares the predicted variatic' of system para ~eters with various trip and systen initia-tion setpoints. The EICSB reviewer evaluates autoratic initiation, actuation delays, possible typas, rodes, interlocks, and the fea sibilf tv of ranual operation if the SAR states that operation action is needed or expected.
To the extent deered necessary, the RSB reviewer enluates tie ef fect of single active fail-ures of systens and corponents which rtay alter tne course of tce transient. This phase of the review uses the systen review procedJrfs described in the ste. d re %w plans for Chapters 5, 6, 7, and 8 of the SAR.
The rathematical nodels used by the appli e t to eveluate core performance and to predict systen pressure in t"e reactcr coola w sys, and main stea n ;ines are reviewed by the RSB to determire if these rodels have been previously reviewed and found acceptable by the staff.
If not, CPB is requcsted tc initiate a aer ic review of the applicant's proposed model.
The values of systen para eters ano ic.itial core and system conditions used as input to the nodel are reviewed by RSB.
Of particular irportar.ce are the reactivity coefficients anJ control rod worths used by the applicant in his analysis, and the variation of rroderator terperature, 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 nininum nargins is evaluated. CFB is consulted regarding the values of the reactivity parareters used in the applicant's analysis.
The results of the analysis are reviewed and corpared to the acceptanc.e criteria presented in Sectiun II of this SRP regarding the maximum pressure in the reactor coolant and main steam,cters.
fhe variations with tire during the transient of paraneters listed in Sections 15.X.X.3(C) ar.d 15.X.X.4(C) of th! StanGard Forrat (Ref. 1) are reviewed. The more irportant of these para ~eters or N,e loss of reactor coolant flow transients are compared to those predicted for other similar plants to verify that they are within the expected range.
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15.3.1-4
Th" revies 2r confirns that a ccrni trent he txen rade in the SAR to ccndJct preoperdtionll tests to verity flou coastdo.vn calculations.
IV.
EVAU!ATION FINDIN'i5 The revie<,er verifies that the SAR contains sufficient information and his review supports the follcwing kinds of statecents and conclusions which should be included in the staff's safety esalurico report (SER):
"Several types of plant occurrences can result in an unplanned decrease in re3ctar coolant ficw rate. The ones expected to occur during the life of the plant are those caused by reactor coolant (or recirculation) prp trips or a flow controller ol'unc-tion.*
All these postul!
- ransients have been reviewed. It was found that the most limiti q in regard to e W r-al m rgins and pressure within the reactor coolant and min stea syste~s was tre _ _
tr3nsient. This transient was evaluated by the applicant using a nathematical model that has t4en reviewed and fourid acceptable by the staff. The values of the para ~eters used as input to this ocdel were reviewed and found to be suitably conservative. The results of the analysis of the transient shon d th3t cladding integrity was maintained by w uring that the mininum departure fror nucleate boiling ratio (DNBR)** did not decrease below and that the maximum pressure within the reactor coolant and nain steam systems did not exceed 110: nf the design pressures "Tne staff concludes that the plant design i' acceptable with regard to transients that are expected to occur during plant life and resu't in a decrease in reactor coolant flow rato V.
REFEPENCES 1.
Regulatory Guide 1.70," Standard Fcrmat and Content of Safe;y Analysis -aports for Nuclear Powor Plants, Revision 2.
2.
ASME Boiler and Preasure Vessel Code,Section III, " Nuclear Power Plant Components," Article NB-7000, " Protection Against Overprcssure," fcerican Society of Mechanical Engineers.
3.
" Standard Safety Analysis Report - BWR/6," General Electric Conpany, April 1973 (under re-view.)
4.
" Reference Safety Analysis Peport - RESAR-3,' Westinghouse Nuclear Energy Systens, November 1973; and " Reference Safety Analysis Report - PESAR-41," Westinghouse Nuclear Energy Sys-tens, December, 1973 (under review).
5.
" System 83 i>andarj Safety AnaT ysis Reenrt (CESSAR)," Cor bustion Engineering, Inc., August 1973 (under review).
- The CSP should prosent nno statenent for all sinilar transients.
15.3.1-5 r 9
),
fi.
" Standard Nuclear Steam System B-SAR-241," Babcock & Wilcox Ccopany, February l'174 (under review).
7.
ANSI N18.2, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," /rerican National Standards Institute (1974).
8.
ANS Trial Use Standard N212, " Nuclear Safety Criteria for the Design of Stationary Boiling Water Reactor Plants," Icerican Nuclear Society (1974).
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