ML19221B142
| ML19221B142 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| 15.02.03, 15.02.04, 15.02.05, NUREG-75-087, NUREG-75-087-15.2.1, NUREG-75-87, NUREG-75-87 15.2.1, SRP-15.02.01, SRP-SRP-15.02.01, NUDOCS 7907120508 | |
| Download: ML19221B142 (6) | |
Text
N U R EG-75/087
[&glha REG Ap hhh,k U.S. NUCLEAR REGULATORY COMMISSION
'j STANDARD REVIEW PLAN N '.U,/'
OFFICE OF NUCLEAR REACTOR REGULATION SECTION 15.2.1 LOSS OF EXTERNAL LOAD, TURSINE TRIP, 15.2.2 LOSS OF CONDENSER VACUUM, CLOSURE OF 15.2.3 MAIN STEAM ISCLATION VALVE (BWR),
15.2.4 AND STEAM PFESSURE REGULATOR 15.2.5 rAILURE (CLC"ED)
REVIEW RESPONSIBILITIES Primary - Reactor Systems Branch (R$B)
Secondary - Ccre Performance Eranch (CFB)
Electrical, Instrumentation and Control Systems Branch (EICSS)
I.
AREAS OF RE'!I_E_W A n#ber of transients wnich are expecd to occur with noderate frequency result in an ur planned decrease in heat removal by the secondary systen. These are covered in this review plan. Each of the transier should be discussed in individual sections of the safety analysis report :SAR), as required by the Standard Format (Ref. 1).
The transients to be evaluated are.
1.
Loss of External Load In a loss of external loa. event an electrical disturuance causem loss of a significant portion of the generator load. Inis loss of load situation is dif ferent f rom the loss of a-c power condition considered in Standard Peview Plan (SRP) 15.2.6 in that offsite a-c power remairs available to operate the station auxiliaries (such as r.. actor coolant pumps). The onsite emergency diesels are therefore not required for the loss of external load transient. Irrediate fast closure of the turbine control valves (ICV) and inter-cept valves is initiated whenever a loss of generator load takes place. For a boiling water reactor (BWR), a fast TCV closure (0.150-0.2 sec) causes a sudden reduction in steam flow and results in a reactor pressure surge. For a BWR without select rod i n';e r t (SRI), reactor scram occurs. For a pessurized rater reactor (PWR) there is also a sudden reduction in steam flow, and this causes the pressure and temperature in the shell side of the steam generator to increase. The latter effect, in turn, results in an increase in reactor coolant tmperature a decrease in coolant density, an increase s
in water volume in the pressurizer, and an increase in reactor coolant pressure. For a PWR with an integrated control system, reactor power can be run back to a lower level on TCV closure.
/ 90i 120 5 og r-7d In all light water-cooled reactors, sensible and decay heat can be renoved through actuation of one or several of the following systems: steam relief system, steam USNRC STAND ARD REVIEW PLAN Svende,d review piene oro prepared foe the goedence of the of%ce of Nucteer Reector Reguietion etoff responsible for the <ev.ew of oppucetione to construct and powee poente These documente ere enede owei6able to the pubhc es port of the Commeseson o polsc, to inform the nocteer industry and the operate nucleet geneest pubhc of regutatory proceducee end pohcies Stendard eeeew piene are not subertutes ter regu etory gedes or the Commiseson a regulatiorie end s
cornpe.ence with thorn re not ei gWred The etenderd rev**w P4en sectione ese heyed to Rev eeon 2 of the Standard Formet and Content of Safety Anetveas Reecets for Nucteet Powee Pfente Not est sections of the Stendeed Formet have e coevooponding review pien Pubhehed standeed rewow plane will be rewtood periodecolly es oppropriate to accommcdote commente and to reviect new m% emet 6on end emettence Commeate end suggese6ene fee impeowement =dt be considered and should be une to the U S Nocteer Reguistory Commession Offeco of Nuciose Reacto' Regulettert Wreshengton. D C 20666 l
bypass to the condenser, reactor core isolation cooling system (RWR), emergency cc cooling systems, and auxiliary feedwater systen (PWR).
2.
Turbine Trip In a turbine trip event a malfunction of a turbine or reactor systen causes the turbine to be tripped off the line by abruptly stopping steam flow to the turbine.
This is dif ferent f rom the loss of electrical load condition described above in that fast closure of the turbine stop valves (TSV) is initiated. The ISV have f aster (0.1 sec) closure times than the tr-bine cont ol val /es, resulting in nore severe transients.
For typical BWR and PWR plants, position swit.hes on the TSV sense the trip and initiate reactor scran The remainder of this transient is similar to the previously discussed loss of electrical load.
3.
Loss of Condenser Vacuun A loss of condenser vacuum event is one of the malfunctions that can cause a turbine t ri p.
The cemarks in 2, above, thus apply to this transient.
4.
Main Steam Isolatinn Valve Closure The main stean isolation valve (MSIV) transient for BWR's can be initiated by verious steam line or reactor system ralfunctions and by various operator actions. As the MSIV's close, position switches initiate a reactor scran when the valves in three or nore of the stean lines are less than 90s onen, the reactor pressure is above 600 psi, and the reactor mode switch is in the RUN position. The effect of MSIV closure is to limit stean flow to the turbine. The results are similar to those discussed in 1, above, but tend to be less severe since the MSIV closure tire is nuch longer than that of the TCV.
5.
Stean Pressure Regulator Failure Steam pressure regulator failure in a closed position yields a transient similar to the transients discussed above. Generally, because the rate of change of systen parameters is slower for a steam pressure regulator failure, a less severe transient results.
The review of the transients described above includes the sequence cf events, the analytical models, the values of parameters used in the analytical nodels, and the predicted consequences of the transients.
The sequ3nt? of events described in the SAR is revieweu by both RSB and EICSB. The RSB reviewer :oncentrates on the need for the reactor protection system, the engineered saf ety systems, and operator action to secure and naintain the reactor in a safe condition. The EICSB reviewer concentrates on the instrumentation and controls aspects of the sequences descr! bed in the SAR to eva.uate whether the reactor and
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plant protectiv:i and safeguards controls and instrunentation systems will function as assuced in the safety analysis with regard to automatic actuation, renote sensing,
indication, control, and interlocks with auxiliary or shared syst s.
EICSB also evaluates potential 'ypass redes end the possibility of manual control by the operator.
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The analytica: methods are reviewed by RSB to ascertain whether the mathematical nodeling and corputer codes have been previously reviewed and accooted by the staff.
If a referenced analytical method has not been previously reviewed, the reviewer requests initiation of a generic evaluation of the new analytical nodel by CPE.
In addition, the values of all the parameters used in the new analytical model, including the initial conditions of the core and system, are reviewed.
The predicted results of the transient analyses are reviewed to assure that the conse-quences meet the acceptance criteria given in Section II, below. Further, the results of the analyses are reviewed to ascertain that the v2. lues of pertinent systen param-eters are within expected ranges for the type and class of reactor under review.
II.
ACCEPTANCE CRITERIA 1.
The basic objectives of the review of the transients listed in Section I are:
a.
To identify which of the moderate-frequency transients that result in an unplanned decrease in secondary system heat renoval are thc most limiting. (The term
" moderate f requency" is used in this review plan in the same sense as in the def-initions of design and plant process conditions in References 8 and 9.)
b.
To verify that, for the most lim ting transients, the plant responds to the transients i
in such a way that the criteria regarding fuel damage anl system pressure are met.
2.
The specific criteria for incidents of moderate f requency '
a.
Pressure in the reactor coolant and main s au system fld be raintained below 110: of the design pressures (Ref. 2).
b.
Fuel cladding integrity should be raintained by ensuring that acceptan:e criterion 1 of SRP 4.4 is satisfied throughout the transient.
c.
An incident of moderate freque cy should not generate a more serious plant condition without other faults occurring independently, d.
An incident of moderate frequency in combination with any single active component f ailure, 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 cladding perforations is acceptable.
3.
The applicant should analyze these transients using an acceptable analytical model.
The eouations, sensitivity studies, and modJls described in References 3 through 6 are
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acceptable. If other analytical methods are proposed by the applicant, these methods are evaluated by the staff for acceptability. For new generic aethods, the reviewer requests an ev luation by CPB.
The values et the parameters used in the analytical model should be suitably conservative.
The fnllowing values are considered acceptable for use in the model:
The reactor is initially at 102% of the rated (licensed) core thermal p^wer (to a.
account for a 21 power reasurement uncertainty).
b.
Conservative scram characteristics are assumed, i.e., maximum tire delay with the most reactive rod held out of the core.
c.
The core burnup is selected to yield the most limiting combination of noderator temperature coef ficient, void coefficient, Doppler coefficient, axial power profile, and ra.ial puwer distribution.
III. REVIEW PROCEDURES The procedures below are used during both the construction permit (CP) and operating license (CL) 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 should be used in the analysis, and the reviewer should compare these to the limiting safety system settings included in the proposed technical specifications.
The description of these transients presented Dy the applicant in the SAR is reviewed by RSB regarding the occurrences leading to the initiating esent. The sequence of events from initiation until a stabilized condition is reached is reviewed to ascertain:
1.
The extent to which normally operating plant instrumentation and controls are assumed to function.
2.
The extent to which plant and reactor protecticn systems are required to function.
3.
The credit taken for the functioning of normally operating plant systens.
4.
The operation of engineered safety systems that is required.
5.
The extent to which operator actions are required.
If the SAR states that any one vf these transients is not as limiting as some other similar transient, the reviewer evaluates the justification presented by the applicant. The applicant is to present a quantitative analysis in the SAR of the reduction-of-heat-removal transient that is determined to be most limiting. For this transient, the RSB reviewer, with the aid of the EICSB reviewer, reviews C 3 timing of the initiation of those protection, engineered safety, and other systems need-to adequately limit the consequences of the transient to an acceptable level. The RSB reviewer compares the predicted variation of system parameters The EICSB reviewer evaluates automatic } [p h 2 b O with various trip and system initiation setroints.
15.2.1-4
initiation, actuation delays, possible bypass modes, interlocks, and the feasibility of manual operation if the SAR states that operator action is needed or expected.
Io th: extent deemed necessary, the RSB reviewer evaluates the effect of single acti'.e failt.res of systems and components which may affect the course of the transient. This phase of the review u W the system review procedures described in the ctandard review plans for Chapters 5, 6, 7, and 8 of the SAR.
The mathematical mdels used by the applicant to evaluate core performance and to predict system pressure in the reactor coolant system and main steam line are reviewed by RSB to determine if these mndels 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 the applicant.
The results of the analysis are reviewed and compared to the acceptance criteria presented in Section II of this SRP regarding the maximum pressure in the reactor coolant and main steam system' The variation with time of parameters listed in Sections.5.X.X.3(c) and 15.X.X.4(c) of the Standard Format (Ref. 1) are reviewed. The mere important of these parameters for the limiting transient are conpared to thnse predicted for other similar plants to verify that they are within expected range.
IV.
EVALUATION FIN"INGS The reviewer verifies that the SAR contains suffic5
- information and his review supports the followin3 kinds of statements and conclusions,..ich should be included in the staff's safety evaluation report (SER):
"A number of plant transients can result in an unplanned decreuse in heat renoval by the sec 21dary system. Those that might be expected to occue with noderate frequency are tt.cbine trip, loss of external 'oad, stean pressure regulator mal functions, main steam isolation valve closure (in BWR's), loss of condenser vacuum, loss of non-emergency a-c power to the station auxiliaries, and loss of normal feedwater flow.* All these postulated transients have been reviewed. It was found that the most limiting in regard to core thermal margins and pressure within the reactor coolant and main steam systens was the transient. This transient was evaluated by the applicant using a mathematical model that had been previously reviewed and fnund acceptable by the staff. The parameters used as input to thic model were reviewed and fauna to be suitably conservative. The results of the analysis of the
. _ _ transient showed that cladding integrity was maintained by cosuring that the maxinum departure from nucleate boiling ratio (or minimun critical heat ratio for a BWR) did not decrease below and that the maximum pressure within the reactor coolont and main steam systems did not exceed 1101 of their design pressures.
"The staff concludes that the plant design is acceptable with regard to transients resulting in an unplannM decrease in heat removal by the,econdary system that are expected to occur witi
'te frequency."
- TIIe SER shouTd present one statement for noderate frequen:y transients involving an unplanned decrease in heat removal by the secondary systen. Thus, the results of the reviews under SRP 15.2.6 and 15.2.7 are included in this statement.
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V.
R tTERE NC E S 1,
Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," Rcvision 2.
2.
A5ME Boiler and Pressura Vessel Code,Section III ' Nuclear PJwer Plant Corponents,"
Article NB-7000, " Protection Against Overpressure," Arerican Society of Mechanical Engineers.
3.
" Standard Safety Analysis Report - BWR/6," General Electric Company, April 1973 (under review).
4 "Referoce Saf"ty Analysis Report - RESAR-7,' Wostinghouse Nuclear Ene gy Systen.-
November 1973; and " Reference Safety Analysis Mcport - RLSAP.-41," Westinghouse Nuclear Energy Systens, December 1973 (under review).
5.
" System 00 Standard Safety Analysis Report (CESSAR)," Combustion Engineering, Inc.,
August 1973 (under review).
6.
" Standard Nuclear Steam System B-SAR-241," Babcock & Wilcox Company, February 1974 (under review).
7.
Standard Review Plan 4.4, " Thermal and Hydraulic Design. '
8.
ANSI N18.2, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," /cerican National Standards Institute (1974).
9.
ANS Trial Use Standard N212, " Nuclear Safety Criteria for the Design of Stationary Boiling Water Reactor Plants," Anericaa Nuclear Society (1974).
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