ML19221B136
| ML19221B136 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| 15.01.02, 15.01.03, 15.01.04, NUREG-75-087, NUREG-75-087-15.1.1, NUREG-75-87, NUREG-75-87 15.1.1, SRP-15.01.01, SRP-SRP-15.01.01, NUDOCS 7907120488 | |
| Download: ML19221B136 (6) | |
Text
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U.S. NUCLEAR REGULATORY CGMMISSION
%$h STANDARD REV EW PLAN
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OFFICE OF NUCLEAR REACTOR REGULATION SECTIC'. 15.1.1 DECPEASE IN FEECWATER TEMPERATP E, INCREASE IN FEE 0 WATER 15.1.2
- FLOW, l'. CREASE IN STEAM rLOW, AND INADVERTENT OPENING OF 15.1.3 A STEAM GE'ERATOR RELIEF OR SAFETY VALVE 15.1.4 REVIEW RESPON4BILITIES Primary - Reactor Soters Branch (RSB)
Secondary - Core Ferformance Branch (CFE)
Electrice!, Instrumentation and Control Systems Branch (EICS3)
I.
AREAS OF REVIEW A number of transients which are ocected to c: cur wi'h roderate frequency, and which involve an unplanned increase in heat renoval by the secondary system, are covered by this review plan. Excessive heat removal, i.e., a heat renoval rate in excess of the he3t r;eneration rate in the core, cause; a decrease in roderator te perature which increases core re3ctivity and can lead to a power level increase and a decrease in shutdown -argin.
The power level increase will lead to a reactor trip. Any unplanred power level increase may result in fuel d3. mage or excessive reactor syste pressure.
Each of the transients covered by tnis plan should be discussed in individual secticns of the 9
safety analysis report (SAR), as required by 'he Standard Forrat (Pef.1). The transients to be evaluated include:
1 f rossurized Water Peactors (PW3'shnd Boiling _ Water _ Peactors (BWR's )
3.
Feedaater syste-alfunctions that result in a decrease in feedwater te pera-ture.
b.
Feedwater system malfunctions that result in an increase in feedw3ter flo.,
c.
Steam pressure regulator ralfunctions or 'aih.res that result in increased steam flow.
2.
PWR's Only a.
Inadvertent o,^ening of a stean generatc relief or,afety valve.
The topics covered in the primary review include: postulated initial core ard reactor conditions which are pertinent to feedwater system ralfunctions, pressure regulator or USNRC STAND ARD REVIEW PLAN Siendard eview piene are preoered for the guidance of the OHice of Nucieer Reactor RegWetson sted 'esponwee f or the rew ow of epo6cetions to construct and operees nucteer power p! ante These documente are eneoe ove lebte to the public ee pad of the Commission e popcy to enform the eucteer Industry and the 6
generos pubhc of reguistory procedureo end pohcies Stenderd review piene are not evbetreutes for reguietory guideo or the Con.misemn a regu6atione end complien:o woh them se nnt requered The etenderd review pian sectione ere bered to Rev;oion 2 of the Stenderd Format end C ontent of Safety Anotveio Reporte for Nucle er Power Ptento Not est sectione of the Stenderd Format have e correspondmg review pien Pub".ned stenoord rev'ew pieno w+H be revised perto$ceHv. se oppropr+ete to accommodate commente end to rehect new mformetson end emportence Commente and suggeenone f or improvemeat will be considered end show d be sent to the u s Nuc#eer R*euiatory Comm.se on Office of Nac#ser Reactor e
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safety or relief valve mlfunctions, methods of therral and hydraulic analysis, postulated sequence of events including tire delays prior to and af ter protective systen actuation, assumed reactions of reactor systen componen'.s functicn11 and operational characteristics of the reactor protection systen in terns of how it affects the sequence of events, and all operator actions required to secure and naintain the reactor in a safe condition.
The results of the transient analysis are reviewed to ensure that the values of pertinent systen pcrameters are within the ranges expected for the type and class of reactor under review. The parameters include: peak clad temperature, Leak fuel temperature, core fles and flow dist-ibution, channel heat flux (average and hot), ninirum critical heat flux ratio (MCHFR) or minimun critical power ratio (MCPR), departure from nucleate boiling ratio (DNER),
vessel water level, thermal power, vessel pressure, steam line pressure (for EWR's), steam line flow (for EWR's), feedwater flos (for EWR's), and reactivity.
The sequence of events described in the SAR for these transients is reviewed by both RSB and EICSB. The RSB reviewer concentrates on the need for the reactor protection system, the engineered safety sysa.:s, and operator action to secure and maintain the reactor in a safe condition. The EICSB reviewer concentrates on the instrument and controls aspects of the sequence described
.'n the SAR to evaluate whether the reactor and pla,it protection and safegaards controls and instrumentation systems will function as assumed in the safety analysis with regard to automatic actuation, remote sensing, indication, control, and inter-locks with auxiliary or shared systems. EIC5B also evaluates potential bypass modes and the possibility of nanual cortrol by the operator.
O The analytical methods are reviewed by RSB to ascertain whether rathematical modeling and co puter codes have been previously reviewed and accepted by the staff.
If a referenced analytical rethod has not been previously reviewed, the reviewer requests initiation of a ceneric evaluation of the new analytical r odel by CPB In addition, the i.alues of all the parameters used in the new analytical model, including the initial conditions of the core and systtn, are reviewed.
II.
ACCEPTANCE CRITERIA 1.
The basic ebjectives of the review cf the transients which result frcm an increase in heat removal are:
a.
To identify which of the roderate-frequency
- transients that result in increaeed heat removal are the rast liniting.
b.
To verify that, for the nost liniting transients, the plant responds to the tran-sients in such a way that the criteria regarding fuel damage and systen pressure are ret.
2.
The specific criteria for incidents of modecate frequency are:
- lhe tern "rodera te-f requency" is used in this review plan in the same sense as in the descriptions of design and plant erocess conditions in References 8 and 9.
)hh 6[j 15.1.1-2
a.
Pressures in the reactor coolant and main steam syste"is should be maintained below 110; of the design values (Ref 2).
b.
Fuel cladding integrity should be maintained by ensuring that acceptance critcrion 1 of Stard'rd Review Plan (SRP) 4.4 is satisfied throughout the transient.
c.
An incident of moderate frequency shoul' not generate a more serious plant condition without other faults occurring irdependently, d.
An incident of moderate frequency in combination with any single active component failure, or single operator error, should not result in loss of function of any barrier other than the fuel cladding. A limited number of fuel rod clidding per-forations is acceptable.
3.
The applicant's analysis of transients caused by excessive heat removal should be per-formed using an acceptable analytical model. The equations, sensitivity studies, and models described in References 3 through 6 are acceptable. If other analytical methods are proposed by the applicant, these methods are evaluated by the staff for acceptability.
For new generic methods, the reviewer requests an evaluation by CPB.
The values of the paraneters used in the analytical model should be suitably conservative.
The use of the following values is considered acceptable:
a.
The reactor is initially at 102 t of the.uted (licensed) core thermal power (to account for a 2 power measurement un ertainty).
b.
Conservative scram characteristics are assumed, i.e., maximum tine delay with the nost reactive rod held out of the core.
c.
The core burnup is selected to yield the most limiting combination of noderator temperature coefficir, void coef ficient, Doppler coefficient, axial power profile, and radial power distribution.
Ill. REVIEW PROCEDURES The procedures below are used for both the construction permit (CP) and operating license (OL) reviews. During the CP review the values of system parameters and setpoints used in the analysis will be preliminary in nature and subject to change. At the OL review stage, final values are used in the analysis and the reviewer should compare these to the limiting safety system settings included in the proposed technical specifications.
RSB reviews the applicant's description of the transients caused by excessive heat removal with specific attention to the occurrences that lead to the initiating event. The sequence of events from initiation until a stabilized condition is reached is reviewed to ascertain:
8 1.
The extent to which nomally operating pl6nt instrumentation and controls are assumed to function.
15.1.1-3
2.
The extent to which plant and reactor protection systems are required to function.
3.
The credit taken for the functioning of normally operating plant systems.
4.
The operation of orgineered saf Hy systens that is required.
5.
The extent to which operatcr actions are required.
If the SAR states that a particular transient involving an increase in heat removal is not as lin1 ting as some other sinilar transient, the reviewer evalu3tes the justification presented by the applicant. The opplicant is tc present a qJantitative analysis in the SAR of the increase-in-heat-removal transient that is determined to be most lir ' ting. For this transient, the RSB reviewer, with the aid of the EICSB reviewer, reviews the timing of the initiation of those protection, engineered safets, and oth r systems needed to limit thc consequences of the transient to an acceptable level. The RSB reviewer conpares the predicted variation of systen parameters with various trip and systen initiaticn setpoints. The EICSB reviewer evaluates automatic initiation, actuation delays, bypass nodes, interlocks, and the feasibility of Tanual operation if the SAR states that operator action is needed or expected.
To the extent deemed necessary, the RSB reviewer evaluates the effect of single active failures of systens and components which nay affect the course of the transient. This phase of tne review uses the system revies procedures described in the SRP for Chapters 5, C, 7 and 8 of the SAR.
The nathe"atical "'odels used by the applicant to evaluate core perfornance and to predict system pressure in the reactor coolant system and nain ste3m lines are reviewed by RSB to deta mine if these nodels have been previously reviewed and found acceptable by the staff.
If not, CPB is requested to initiate a generic review of the model proposed by thc applicar.t The values of system parameters and initid! core and system conditions used as input to the model are reviewed by RSB. Of particular importance are the values of reactivity co-efficients a.d control rod norths used by the applicant in his analysis, and the variations of noderator t eperature, void, and Doppler coef ficients of reac tivity with core life. The reviener evaluates the justification provided by the applic3nt to show that the core burnup selected yields the mininun nargins. CPB is consulted reg 3rding the values of the reacti'/ity parameters used in the applicant's analysis.
The results of the analysis are reviewed and comoared to the acceptance criteria presented in in Section II of this SRP regarding the maximJ:, pressure in the reactor Coolant ar i nain stea, systems. The variatico with time during the transient of paraneters listed in Soctions I S.x.x. 3(c) and 15.v.x.4(c ) of the Standard Forna t (Ref. 1) is reviewed. T h. valuos of the note important of these para eters, as listed in Section I of this SpP, are co m r H to those predicted for other similar plants to see that they are within the rar y expected.
15.1.1-4 7, n l / n/
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IV.
EVALUATION FINDINGS The reviewer verifies that the SAR contains sufficient informatico and nis review supports the following kinds of state"'ents and conclusions, which should be included in the staff's safety evaluation report:
"A nunter of plant transients can result in an unplanned increase in heat removal by the secondary system. Those that night be expected to occur with noderate frequency can be caused by feedwater syster er pressure regulater malfunctions or the inadvertent opening of a stea"' cenerator safety or relief /alve (WR only). All of these postu-lated transients have been reviewed. It was found that the nost limiting in regard to core thermal nargins and pressure within the reactor coalant and rain stean systers was the transient. This transient was tvaluated by the applicant using a nathematical model that had been previously reviewed and found acceptable by the staff.
The para";eters used as input to this model were reviewed and found to be suitably con-servative. The results of the analysis of the transient showed that cladding integrity was maintained by ensuring that the ninimum departure from nucleate boiling ratio (D:,BR)* did not decrease below and that the maxirun pressure within the reactor coolant and main steam systers did not exceed 110' of the design pressures.
"Tne staf f concludu + hat the plant design is acceptable with respect to transients resulting in an unplanned increase in heat removal by the secondary systen that are expected to occur with moderate frequency.'
V.
_PEFERENCES 1.
Pegulatory Guide 1.70, ' Standard Format and Ccntent of Safety Analysis Reports for Nuclear Power Plants,' Revision 2.
2.
ASME Boiler and Pressure kessel Code,Section III, ' Nuclear Pcwor Plant Components, Article NB-7000, " Protection Against Overpressure," American Scciety of Mechanical Engineers.
3.
" Standard Safety Analysis Peport - BWR/6,' General Electric Conpany, April 1973 (under review).
J.
" Reference Safety Analysis Report - RESAR-3,' Westinghouse Nuclear Energ/ Systems, November 1973; and " Reference Safety Analysis Report - RESAR-41," Westinghouse Nuclear Energy Systers, December 1973 (under review).
5.
" System 80 Standard Safety Analysis Report (CESSAR),' Combustion Engineering, Inc.,
August 1973 (under review).
6.
" Standard Nuclear Steam System B-SAR-241," Babcock & Wilcox Corpany, February 1974 (under review).
7.
Standard Review Plan 4.4, "Thernal and Hydraulic Cesign."
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15.1.1-5
8.
ANSI N18.E "fuclear Safety Criteria for the Dr. sign of Stationary Pressurized Water Reactor Flants,' A~erican flational Standards Institute (1974).
9.
ANS Trial Use Standard N212, " Nuclear Safety Criteria for the Cesign of Stationary Boiling Water Reactar Plants, American Nuclear Society (1974).
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15.1.1-6
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