ML19221B158
| ML19221B158 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| 15.04.05, NUREG-75-087, NUREG-75-087-15.4.4, NUREG-75-87, NUREG-75-87 15.4.4, SRP-15.04.04, SRP-SRP-15.04.04, NUDOCS 7907120544 | |
| Download: ML19221B158 (5) | |
Text
NUREG 75/087 fgaAg'%
,M U.S. NUCLEAR REGULATORY COMMISSION
%eV@) STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION e
SECTION 15 4.4 STARTUP OF AN IMCTIVE LOOP OR RECIRCULAT;0N 15.4.;
LOOP AT AN INCCRRECT TEWERAfURE, %D EL0i' CONTROLt ER MALFUNCTION CAUSIN'i AN INCRE ASE IN BWR CORF flnW RAT [
R E V I_E W R E.Srf.S.I B_I L I T I E.S Primary Ceactor Systens Branch Secondary - Electrical Instr, entation and Control Systems Branch (EICSB)
Core Perfonunc< Branch (CP3)
I.
A R.F A S O F _K E V I.E W-A nu ' er of transient. that "ay occur witn noderate f rmency cause eitner introased core f i c es or introiuction of cooler or de-borated water into the care.
Th"se transiente result in an increase in cr re reactivity due to decreasm) moderator te ;mrature, moderator boron concentration, or <cre void fraction. This review plan is intered to t;e applicable to all such transion'=
Each of tFese transitnts shuuld be discussed in individal sectia; nt the applicant's safets analysis report /wi is required t)/ the Standard Fcrrit
/ Pef. 1).
Tne specific transients (Table 15-1 cf Mf. 1) evaluitod are:
1.
R3ilim water rec tor (EWR): startup of an idle recirculation rump.
2 BWR: flow cont-aller malfunction causing increased recirc h ion flew.
3.
Pressurized water reactor (PWR) with loop isolation valves, sty tup of a ruv in an initially isolate 1 inictive reactor coolant loop whero the rate of flow increase is linited by the rate at which the isolation valves open.
4.
PWR withoJt locp isolation valves: startup of a p rp in an inattive loop.
Th< review of the core flow increase transients considers the sequen:e of emnts, the anal /tical rodel, the values of para ~eters used in the anilytical r odel, and the predicted conser,uences of the tran',ient.
7907120 W 4 USNRC STAND AR7 RE\\r!EW PLAN st.nd.,d eev.,6ene e.. p..pe..d e, th. g uidenc e o.he owe of N ucies, Reecio. aieguise,on e et+ eesponsibee v r the r..... oi.ppu,.etions to ro,,si,uct on d e
o operaie nuca.or po.., pionee These documeate see made evedobie to the pub er ao port of the Commese+on e polic y to 6nf orm the nuciose industry and the genered pubhc of requietory procedures and potec.ee standerd rev ew plane are not oubat tutes f oe rogue cry guides or the Commseason a requietione encc compliance w*th them se not requered The etendeed review pian sectione are beved to Revision 2 of the Standard f ormer and Conteet of Se*ety Aner e a Napoets v
for soucteer Powse Ptente Not est sect one of the Stenderd Formet have a cortempondmg eeview pien Published stenderd review plane will be revised penodit ouw ao oppropriate to accommodete cornment, and to eetie.:
new,,s
,mes on and esper nc..
o Commente and suggestions for emprovement wen be cone.dered and should be eeni to the U S Nudoes Requistory Commeesion Office of Nucieer Reactor Regenteon Weehengton D C 20bh6
,7
The seq;ence of events described in the SAR is reviewed by botn RSS and EICSB. The PSB revieaer concentrates on the need for the re3ctor protectich syste'" and operator action to secure and aintain tre reactor in a safe condition. The EICSB reviewer concentrates en the instru-entation and controls aspects of the sequences described in the SAR t evaluate whether the re3ctor and plant protection and safequards controls and instru~entation systems will function as ass reJ in the safety analysis with regard to automtic actuaticn, re"ote sensing, indication, control, and interlocks with auxiliary or shared systems.
FICSB also evaluates potential bypass odes and the possibility of man;31 control by the o erator.
The analytical retnods are reviewed by RSS to ascertain whether the Mthematical odeling and cc~puter codes have been previously reviewed and accepted by the staff. If a ref?renced 3nalytical "ethod his not been p eviously reviewed, the revienor requests CPB to 'nitiata a ceneric enlustion of the new analytical odel. In addition, the values of all the parr'eters used in the new analytical nodel, including the initial conditions cf the core and systr, are rovieaed.
Tne predicted resalte of the transients are reviewed to assure that the consequences ~eet the acceptance criteria given in Section II, belos.
Further, the results of the transients are reviewed to ascertain that the values of pertinent systen carameters are within ranles e cected for the tr a and class of reactor under revies.
II.
ACCE?iu.CE CRITERIA 1.
The basic cbjectives of the review of the transients described abase are:
a.
To identif/ wh'ch of the transients are the nost limiting.
b.
To verify that, for the most limiting transients, the plant responds in such 3 W3y th3t the criteria reg 3rding fuel da930e and systen pressure are
'et.
2.
The soecific criteria for incidents of moderate frequency are:
a.
Pressure in the reactor coolant and nain stean systems should be naintained below 110~ of the design pressures (Ref. 2).
b.
Fuel clad integrity should be maintained by ensuring that acceptance criterion 1 of Standard Review Flan (5RP) 4.4 is satisfied throughout the transient.
c.
An incident of moderate frequency should no*. generate a nore serious plant condition without other faults occurring independently.
d.
An incident of noderate frenuency in corbination with any single active component failure, cr s ngle operator error, should not cause loss of function of any barrier, other than the fuel cladding, to the release of radioactive raterials. A limited number of feel rod cladding perforations is acceptable.
15.4.4-2 Q
~,5 }
f i
3.
The applicant's analysis of the nost limiting transients should be performed using an acceptable an31ytical rodel. The equations, sensitivity studies, and rodels described in References 3 through 6 are acceptable. If other analytical rethods are proposed by the acplicant, they are evaluated by the staff for acceptability.
For new generic rethods, the reviewer requests an evaluation by CFB.
The values of the parameters used in the analytical nodel are to le suitauly con-servative. IFe following values are considered acceptable:
a.
Initial power level is rated output (licensed core ther al power) for the number of loops initially assu'"ed to be operating, plus an allowance of 21 to account for power measurement uncertainty. An analysis to determine the effects of a flow increase must be rade for each allowed mode of operation (i.e., one, two, or three loops initially operating) or the effects referenced to a limiting case.
b.
Conservative scra71 characteristics are assu ed, i.e., maxinu 1 tine delay with the rost reactive rod vield out of the core, c.
The core burnup is selected to yield the Post limiting combination of oderator te"perature coefficient, void coefficient, Doppler coefficient, axial pener orofile, and radial power distrit'u' ion.
III. PEVIEW PROCEDURES The procedures belcw are used during both the construction perm!+ (CP) and coeritir': license (OL) re/iews. During the C? review, the values of systen para eters and setpoints used in the analysis will be preliminary in n3ture and subject to change. At the OL review, final values shou!d be used in the analyses, and the reviewer should co"m re these to the liniting safety settings included in the proposed technical specifications.
The description of the core flow increase transients presented in the SAR is reviesed by 453 regarding the occurrences leading to the initiating event. The sequence of events frcn initiation until a stabilized condition is reached is reviewed to ascertain:
1.
The <xtent to which normally operating plant instra entation and controls are assumeti to functiu'i 2.
The extent to which plant and reactor protection systems are required to function.
3.
The credit taken for the functioning of nor ally operating plarit systems.
) t/)
'54 15.4.4-3
4 The operation of engineered s3fety s/stens that is required.
5.
The extent to which operator actions are required.
If the SAR st&'.e5 that a particular core flos increise transient is rat as liniting as sore other similar transient, the reviewer evaluates the justification presented by the applicant
!*e applicant "hould present a quantitative analysis in the SAR of the increase of flow tran lent that is determined to be no;t limiting. For this transient, the RSB reviewer, sith the aid of the EICSC reviewer, reviews the timing of the ini tiation of protection, engineered safety feature, and other syster s needed to li mit tne consequences of the core floa increaso transient to acceptable levels The RS3 revieser comparos the predicted variation of s/stm para ~eters with various trip setpoints. Tho EICSB revieaer evaluates automatic initiation, actuation delays, possible bypass r, odes, interlocks, and the feasibility of minual operation if the SM states thit Gper stor action is needed or e gected.
To the t tent dee ed necessary the R't3 reviener evaluates the effects of single active failures of systems and components which N / a'fect the course of the transient. This phase of the review uses the syster review procedures described in the standa rd revw plans for Chapters 5.
f>, 7 and 8 of the SAR. The revieaer considers and evaluates the possibility of a sin J e f ailure that would permit the loop isolation valves to open prior l
to startup of a para in an idle loop (for 40se plants with loop isolation valves). If this coulj cccur, the core floa rate increa5e WoJld not be limited by the rate at wnich the valve opens, and the resulting rate of reactivity insertion could be greater than for other transients of this group.
The nathematical nodels used by the applicant to evalu3te core performance and to predict systen pressure in the reactor coolant system and cain steam line 3re rev'ewod by R50 to determine if these mode? s have been previously reviesed and found accepta' le by the st3'f.
If not, CPB is requested to ir.:tiate a generic review o' the nodel propr aed by the applicant.
The values of systen para eters and initial core anj sy>te, corditions used as input to the rodel are revieued by PSC. Of particulde inportance are the reactivity coef ficients and control rod worths used in the applicant's analysis, and the variation of moderator te perature, void, and Doppler coefficients of reactivity with core life. The justification provided by the applicant to show that the selected core burnup yields the nininun nargins is evaluated. CPB is consulted regarding the values of the reactivit/ 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 the naxinum pressure in the reactor coolant and mai.i steam systems. ror each transient the variations with tire during the transient of the core and barrier perfornance paraTeters listed in the " Event Evaluation" section of Ch3pter 15 of the Stindard Format (Ref. 1) are reviewed. The values of the more important of these parameters for the core flow increase transients are compared to those predicted for 7ther similar plants to see that they are within the range expected.
[ I) j [])
15.4.4-4
IV.
EVALUATION FINDINGS The reviewer verifies that the SAR c'ntains sufficient inform 3 tion and his review supports the folloveing kinds of state"cnts and conclusions, which should be included in the st3ff S safet f evaluation report ISEP):
' A nu~ter of plara transients c an result in a core floa increase. Those chat miqat he eqected to occur with moderate freq;ency are the startup of an idle recircJlation [urp (CWR); floa controller ralfunction causing increasing core fica kWR); startup of a purp in an inactive reactor coolant loop (PWR); and startup of a prp in an initially isolated in3ctive reactor coolant loop
- All these postulated transients have beco reviewed.
it was found that the "ost liniting with reg 1rd to core thernal nargins and pressure within the re3ctor coolant and main steam systens was the _
_ transient. This t 'ansient was evaluated cy the applicant using a m3 thematic 3) noc'el that has been previously reviewed and found acceptable by *.he st3ff. The paraneters used as input tc this model were reviewed and found to be suitabl) conservative. Tne r esults of the 3931ysis of the transient sho.ved trat cladding inteqrity was naintained by ensuring that the minima, departure f rom nucleate boiling ratio (MD'.3R)" did not decrease l o"s
- -... ud that the Caxi"'u, pressure within the reactor ccolant and nain stea'i sjste, pressures did not exceed 110 of the design values.
"The staf f ccncludes that the plan: design is acceptable in ret 3rd to transients tLat result in an increase in coolant flow through the reactor core.
9 1.
d @ '.C E S 1.
%:;;latory Guide 1.70, " Stand)rd F]rnat and Centent of $3fety Analysis ?eparts for
'. Alea r Power f lant s, Pevisien 2.
2 M"
Eniler and Pressure Vcscel Code,Section III, Lclear Power Plant Correnents, Art i c le '.5-7000, ' Protect iu. Ag a in st O serormssure, A erican Societ_v of f'echanical Engineers.
3.
" Standard Safete Analysis Percrt - GR /6, ' Gener3! Electric Co~ any, Acril 1973 (ander review).
4.
Reference 53fety Analysis Re p rt PESAR-3, Westin'; house Lclear Energy Sjste s,
' ove"ter 1973; and " Reference Sa fot y An.ilysis Pecert - RESAR-al' (under re/iew).
5.
"5 ste, 80 Stand 3rd Safety Analysis Report (CESSAW), Co.bustic,n Engineering, Inc.,
f August 1973 (under review).
6.
" Standard Nuclear Stea~ Syste' B-SAR-241, Babcock & Wilcox Co pan,, Feb mary 1974 (under reviea) 7.
Standard Revica Plan 4.4, "Thernal and Hydraulic Design.'
\\h b'
- fhe SER snould present one state ent far all similar transients
- The nin1"un critical hedt flux ratio or crit cal poner ratio (MCH R or U P) for a EWR.
15.4.4-5