ML19221B159
| ML19221B159 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-15.8, NUREG-75-87, NUREG-75-87 15.8, SRP-15.08, SRP-SRP-15.08, NUDOCS 7907120545 | |
| Download: ML19221B159 (7) | |
Text
N U R E G-75/087 p W G eq f
sk U.S. NUCLEAR REGULATORY COMMISSION N[' ih' /
OFFICE OF NUCLEAR REACTOR REGULATION i
STANDARD REVIEW PLAN 9
\\ 'N SECTIC'. 15.8 ATICIPATED TRANSIENTS WITM UT SCP W R_E_V I_E W R i S_n_ N_S_I B I_L_I_T_I_E_5-Pririry Reactor Systers branch (RSB) 9 ondiry - Accident Analysis Branch (AAB)
Centain ent Syste s Branch (CF9)
Core Ferformance Branch (CfB)
Electrical, instrumentation and Centrol Syste s Branch (LIC50)
Mechanical Engineering Branch (CES)
I.
AREAS OF PEVILW Anticipated transients are transients expected to occur during the lif e of tN plant Anticipoted transients witnout scra~ ( ATWS) are those low probability events in eich en anticipated transient occurs cod is not followea by an autoratic reactor shutdown (scrar). The f ailure of the reictor to shJt dawn auickly dJring those transients can lead to unacceptable reactor coolant systcc pressures and to fuel da ;ge.
Typical transients that ray have unacceptable com uences if thero ;s a scran f ailure are: loss of f oodwater, ioss of load, turbine trip, inadvertent control rod with1rawal, Ir>> o f a-c pcwo r, less of condonser vacuum, and, in a boiling v3t?r reactor (BWR), closure of main ste rlire isolation valtes. The effects of all ATWS events rust be evaluited, including a corplete ancljsis 9
for those resulting in the rost severe conseqsences fuh should te discussed in indi vi d;il soctions of the safety at:alysis repor, (.
as reaaired by the St3ndard Format (Pef. 1).
k cording ta the licensino oition en ATW5 (Ref. 2), plants for which a construction f or"it (Cf) application is filed af ter Oi tober I, 1976 rust ha se rcactor shutdown syste", of such high reliability that ATWS events need not be considered This stand 3rd review plan (5?P),
tnerefore, applies only to plants,,here applicatien was r ade for a CF ' efore tha t d3 te.
For seth plants, the review of ATW5 events is concerr,ed with the sea,ence of e'.en's, tho analytical rodel, the values of paracters used in the analytical rr, del, and the predicted results and consequences of the transient.
T: seq >ence of emnts described in the applicant's SA? is reviewed by both R5B and EICSB The RSC reviewer concentrates on the effect of the reactor protection system, engineered safety systems, other tsperational and control syste"s. and operator action on the course of the transient and the securing and maintenance of the reactor in a safa con d i ti on. The 790~/1209/
USNRC STANDARD REVIEW PLAN stenoere review piece are proper.d to, the go.dence er the o'+ ce of Nucreer Reecto, anguiet oa ete+f roepon. b4, fo, ihe v.ew of oppu enr.e to construci and c
opetete nucteer power pisets These documen?e are made evadeoie to the puhuc es part af the Comm,se.on a pohey to soform the nucceete,dustry and the 8
genere4 pubhc of regulatory proceduree end pohc ee Stendeed review p8ene are not substitutes for reguietory guidee ce the Commission a eequieteone on; compheace w'th them to not roqua**d The eter.derd review plan sec.tione are heyed to Reves.on 2 of the Standard Formet and Conton'sf Sof,ty Anotys e Reporte foe Nucleo# Power Ptente Not e#i sectioco of the 5'endeed Format beve e coneapoor ng rev.ow pian e
Pubheneu etenderd rewow piens will be revised periodiceity es appropr.ete to accommodate core, ente end to regioct ce u it.f ormet on end e.per once Commente and suggestions for 6mprovement eds be considered and should be eent to the U 5 Nucteer Reguietory Coreguep.s@*f.e e et Nascie, peertor Ref unet on Washington D C 24f6 yn i b8 Y$ MY' EMF )f***
1 50 056
LICSB reviewer concentrates on the instrurentation and controls aspects of the sequence described in the SAR to evaluate whether the reactor and plant protection and operati,,1al safeguards controls and instrumentation systems (other than the scram system which is 9
assured to have failed) will function as assumed in the safety analysis with regard to automatic actuation, remote sensing, indication, control, and interlocks with auxiliary or shared systems. EIC:B also evaluatts potential bypass modes and the possibility of manual control by the operator.
Thc analytical methods 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 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 parar:eters used in the new analytical model, including the initial conditions of the core and system, cre reviewed.
CSG reviews ATWS effects on the containrent syste:' with regard to the ef fects of the result-ing pressures and te peratures on its f unctional capabilities.
MEB reviews the ef fects of ATWS eventi on the structural integrity of the reactor coolant pressure boundary.
AAB evaluates the fission product release assurptions used in determining any of fsite releases and verifies that the radiological consequences resulting f rom ATWS events are within acceptable linits. (See attached appendix.)
the The predicted results of ATWS e;/ents are reviewed to assure that the consequences meet acceptance criteria given in Sc tion II telow. Further, the results of the analyses are reviewed to ascertain that the v1 lues of the pertinent initial conditicns and system para"eters are within ranges expe1ted for the type and class of reac tcr under review.
II.
ACCEPTANCE CRITERIA 1.
The general objectives in the revfew of ATWS events are:
To identify the ATWS events tha' are rost limiting.
3.
b.
To verify that for the rost linitirg events the specific criteria applicable to ATWS events are ret.
2.
The speci fic criteria that apply depend on the date the CP application is filed and wnether the need to take ATWS into account was noted in the staff's safety evaluation report (ELR) or in the report of the Advisory ommittee on Reactor Safeguards (ACRS).
c If the nted to take ATWS into account war
- ed in either of these reports and if an application for a CP is filed before October 176, then the following criteria (4pendix A of Ref. 2) apply:
~
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15.8-2
Calculation of Cons 1quon_c_e_s a.
The calculated radiological consequences should be within the guideline values set forth in 10 CFR Part 100.
In additior, the linits listed below on calculated systee pressure, fuel performance, and containrent conditions should be ret.
3 R e a r t_o r_ _L oo_l a n t_5ys ten P re s s u re The raximum acceptable calculated transient reactor coolant system pressure should be based cn the system boundary pressure limit or the fuel pressure limit, whichever is nere restrictive:
(i!
R_ea c tor, Co_o_l ant _Sys tt m Bounda ry_ Pres sure Lini t The calculated reactor coolant system transient pressure should be linited such that the raxirum primary stress anywhere in the system boundary is less than that of the " emergency conditions" as defined in the ASME Code,Section III (Ref. 3).
(ii) Fuel Pressure Limit The calculated reautor coolant system transient pressure should not exceed a value for which tests and analyses deronstrate that there is no significant safety problen with the fuel.
2.
Fuel Thermal and Hydraulic Performance (i) The calculated average enthalpy of the hottest fuel pellet should not result in significant cladding degradation or significant fuel nelting.
(ii) A calculated critical heat flux event should r.ot occur unless the calculated peak claoding temperature can be shown not to result in significant cladding degradatico.
3.
Containnent Conditions Calculated maximum containment pressure should not exceed the design pres ure of the containment structure. Equipnent located within the contain ent that is relied upon to mitigate the consequences of ATWS should be qualified by testing in the combined pressure, temperature, and hunidity environment conservatively predicted to occur during the course of the event.
b.
Evaluation Techniques Analysis models and techniques, including computer codes, used for conservative evaluations of the consequences of postulated ATWS events, together with assoc-iated assurrptions and parameters, should be described and justified in topical reports.
Review of Reactcr Shutdown System Design c.
A review of the reactor shutdown systen design should be made with the aim of identifying and correcting areas that might be particularly vulnerable to corron mode failures.
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058 rn a
d.
Uiversity_Regutement_s Design changes to make the calculated consequences of postulated ATWS acceptable should not rely on equipnent which has a failure mede comon with the anticipated transient or the shutdown systen. To the extent practical, the equipment in-volved in the design should operate cn c different princiole from equipment in the shutdown systen As a minimum, the equipnent relied on to nake the con-sequences of a postulated ATWS event acceptable should not include equipment identical to equipment in the associated shutdown systen. Such designs should be shown not to result in violations of safaty criteria for steady-state, transient,
or accident conditions and should not adversely af fect the operation of any safety-related systers.
III.
PEVIEW PR_0CEDU_RES The review procedures below are used during both the CP and operating license (UL) reviews for plants for which this plan is applicable. During the LP review the values of systen parreters and setpoints used in the analysis will be preliminary in nature and subject to change. At the OL stage, final values should be used in the analysis and the reviewer should compare these to the expected values, control systen Dands, and limiting safety systen settings included in the proposed technical specifications.
The description of the ATWS events (Table 15-1 of Ref.1) presented by tN applicant in the SAR are reviewed by RSB regarding the occurrences leading to the
.itiatim event. The sequence of events from initiation until a stabilized condition is reacned is reviewed to ascertain:
1.
The extent to which nor ally operating plant instru entation and controls are assu ed to function.
2.
The extent to which engineered safety system actuation instrumentation arid controlc are required to function, and the credit taken for reactor trip alar ~s to inform the operator of abnor al conditions.
3.
The credit tak en for the functioning of normally nreratir,g plant syster s.
4.
The operation of engineered safety systens that is renuired.
5.
The extent to which operator actions are required.
Ihn applicant should present a quantitative analy;is in the SAR of the ATWS events that are deternined to be r.ont liniting. The justification for selecting these events should bo given. The R$B reviewer, with the aid of the EICSB reviewer, reviews the timing of the initiation of this protection, engineered safety, and other systens reeded tc limit the consequences of the events to acceptable levels. The RSB reviewer compares the predicted variation of system parameters with various trip and system initiation setpoints. The EICSB reviewer evaluates auto atic initiation, actuaticn delays, oossible b/ pass rodes,
}I'] ]}h 15.8-4 luu o' s-l r
interlocks, and the feasibility of manual operation where the SAR States that operator action is.eeded or expected. He also verifies tha* the applicant has reviewed the design of the reactor shutdown system and identified and corrected areas which are vulnerable to comnon mode failures. Both the RSB and EICSB reviewers verify that any design changes rade to mitigate the consequences of the ATWS events have no failure rode cerron with the cause of the anticipated transient or with the shutdown system The c athematical c.odels used by the applicant to evaluate core perfonnance and to predic t systen pressure in the reactor coolant system and m.ain steam system are reviewed by RSB to deternine if these models have been previously reviewed and found acceptable by the staff.
If not, CFB is requested to initiate a generic review of the rodel prnposed by the applicant.
The values of system paran9ters and initial core and system conditions ; sed as input to the nodel are reviewed by RSB. Uf particular irportance are the reactivity coef ficients used by the applicant in his analysis and the variation of noderator temperature, void, and Doppler coef ficients of reactivity with core life. The justification provided by the aEplicant to show that he has selected the core burnup that yields the minirun margins is evaluated. CFB is consulted regarding the values of the reactivity parameters used in the applicant's analysis.
The results of the analysis are reviewed and compared to the acceptance criteria presented in Section II of this SRP regarding maximum pressure in the reactor coolant and main steam systers and fuel performance limits. For each ATWS event the variatica with time during the transient of the core and barrier performance parameters listed in the " Event Evaluation" section of Chapter 15 of the Standard Fon"at (Ref. 1) are reviewed. The c. ore in portant of these para eters for ATWS events are compared to thoso predicted for other similar plants to verify that they are within the expected range.
IV.
EVALUATION FINDINGS The reviewer verifies that the SAR contains sufficient information and his review supports the following kinds of staterents and conclusions, which should be included in the staff's safety evaluation repart:
"A nur,ber of plant transients can be af fected by a f ailure of the scram systen to function. For a PWR* the nost important include loss of feedwater, loss of Icad, inadvertent control rod withdrawal, and loss of a-c power. All postulated anticipated transients whicn can be affected by a failure to scram have been
.iewed.**
It was found that the rest limiting ATWS events, with respect to possibi, fuel damage and "For a EAR they are loss of condenser vacuum, closure of the nain steam isolation valves, and turbine trip.
- The SER sheuld present one statorent for all similar transients.
m h
4 i x" 150 000 15.8-5 e
pressure within the reactor coolant and main steam systems, were the
_ and events respectively. The events were evaluated by the applicant using a mathematical model that has been reviewed and found acceptable by the staff. The parameters used as input to this model were reviewed and found to be suitably con-servative. The results of the analysis showed a peak reactor coolant system pressure of psia which is within the ' emergency conditions' defined in Section III of the ASME Boiler and Pressure Vessel Code. The calculated (fuel clad te"'perature, DNBR, CPR, or MCHFR) following a postulated transient without scram is and will not result in significant fuel damage. The radioactivity release from this event has been calculated and the resulting doses found to be within the guidelines of 10 CFR Part 100.
" Based on the above evaluation, the staf f concludes that the plant design is acceptable with regard to anticipated transients without scran. '
V.
R E F E_E EEC E S_
l.
Regulatory Guide 1.70, " Standard Forrat and Content of Saft ty Analysi Reports for Nuclea" Power Plants, Revision 2.
2.
Fegulatory Staff, " Technical Report on Anticipated Transients Without Scrar for Water-Cooled Power Reactors, WASH-1270 U. S. Atomic Energy Comission, Sept. 1973.
3.
ASME Boiler and Pressure Vessel Code,Section III, thclear Power Plant Components,'
Article NB-3000, Design, ' Arerican Society of Mechanical Engineers.
15.8-6 m
N,Y
~v
APPENDIX STANDARD REVIEW PLAN 15.8 RADIOLOGICAL CONSEQUENCES OF AN ATWS EVENT REVIEW RESPCNSIBILITIES Primary - Accident Analysis Branch (AAB)
Secondary - Site Analysis Branch (SAB)
Reactor Systems Branch (RSS) 1.
AREAS OF REVIEW The review of ATWS events will include estimates of the radiological consequences. These estimated consequences will be compared to those estirrated for design basis events as well as the exposure guideline' of 10 CFR Part 100 and Regulatory Guides 1.3 and 1.4.
If the ccnsequence estimates for any ATWS event are significant when co'" pared to those for design basis events or require imposition of a technical specification to limit the conse% ences, then that ATWS event will be included as a design basis event in the staff safety evalua-tion. The review method and plan are under development and will rely on the result of current staff reviews of vendor analyses of ATWS events.
150 062 lE va 15.8-7