ML19256B798
| ML19256B798 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/11/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19256B797 | List: |
| References | |
| TASK-TF, TASK-TMR NUREG-75-087, NUREG-75-087-15.6.1, NUREG-75-87, NUREG-75-87-15.6.1, SRP-15.06.01, NUDOCS 7908310011 | |
| Download: ML19256B798 (8) | |
Text
NUREG.75/067
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STANDARD REVIEW PLAN s
OFFICE OF NUCLEAR REACTOR REGULATION SECTICN 15.5.1 INACVERTENT CPENING OF A PWR PRE 55URIZER SAFETY / RELIEF VALVE CR A BWR SAFETY / RELIEF VALVE
_ lEVIE'd RESPCNSIBILITIES Primary - Reacter Systems Branch (RSB)
Secondary - Core Performance Branch (CPB)
Electrical. Instrumentation and C:ntrol Systems Branch (E!CSB)
I.
AREAS OF REVIEW The inadvertent opening of a safety or relief valve results in a reactor coolant inventory decrease and a decrease in reactor coolant system pressure. The effect of the pressure decrease is to decrease the neutron flux (via moderator density feedback). In a pressurized
.ater reactor (PWR) a reactor trip occurs due to low reactor coolant system (RCS) ;t-ssure.
In a toiling water reactor (BWR) the safety or relief valve discharges into the su;;ression
- ool. Nor ally there is no scram in a BWR. The pressure regulator senses the RCS pressure decrease and partially closes the turbine control valves (TCV) to stabilize the reactor at a
'ower pressure. The resctor power settles out at nearly the initial power level. The coolant inventory is maintained by the feedwater control system using water frem,the c:ndensate storage tank via the condenser hotwell.
The review of these transients should consider the sequence of events, the analytical codel, the values of parameters used in the analytical model, and the predicted c:nsequences of the transient.
- he secuence of events described in the applicant's safety analysis report (SAR) is reviewed by both RSB and EICSB. The RSE reviewer concentrates en the need for the reactor protect. ion system, the engineered safety systems. and coerator action to secure and maintain the reactor in a safe condition. The EICSB reviewer c:ncentrates on the instrumentation and controls asoects of the sequence described in the SAR to evaluate whether the reactor and plant protection and safeguards controls and instrumentaticn systems will function as assumed in the safety analysis with regard to aut:r.atic actuation, reecte sensing, indication. c:ntrol, and interlocks with auxiliary or shared systems. EICSB also evaluates potential bypass modes and the possibility of manual control by the operater.
1909 344 USNRC STAND ARD REVIEW PLAN
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The analytical methods are reviewed by RSB to ascertain whether the athematical modeling and computer codes have been previously reviewed and accepted by the staff. If a referenced method has not been previously reviewed, the reviewer requests initiation of a generic h
evaluaticn of the'new analytical medel by CPB. In additien, the values of all the para-b meters used in the new analytical model, incit. ding the initial conditiens of the core and system, are reviewed.
The ;redicted results of the transient are reviewed to a'ssure that the consequences meet the acceptance criteria given in Section II of this standard review plan (SRP}. Further, the 1
results of the transients are reviewed to ascertain that the values of pertinent system parameters are within ranges expected for the type and class of reactor under review.
II.
ACCSpTACE CRITERIA 1.
The general oojective in the review of inadvertant prieary safety or relief valve opening events is to confirm that one of the following criteria is met:
The consequences of the transient are less severe than the consequences of another a.
transient that results in a decrease of reactor coolart inventory and has the same anticipated frequency classification, b.
The p: ant responds to the safety or relief valve opening transient in such a way that the criteria regarding fuel damage and system pressure are met.
2.
The specific criteria for incidents of rederate frequency are:
E a.
pressure in the reactor coolant and main _ steam systems should be maintained below N+
110% of the design pressures (Ref.1).
b.
Fuel clad integrity snould be maintained by ensuring that acceptance criterion 1 of SRP 4.4 (Ref. 7) is satisfied throughout the transient.
c.
An incident of moderate frecuency sh,uld not generate a more serious plant condition without other faults occurring independently, d.
An incident of moderate frequency in ccmbination with any single active component failure, or single operater error, shculd not cause loss of function of any barrier other than the fuel cladding. A limited number of fuel rod cladding perforations is acceptable.
3.
The applicant's analysis of this transient should be performed usiig an acceptable analytical medel. The e:vattens, sensitivity studies, and models described in References 2 througn 5 are acceptable. If other analytical methods are ; reposed by the applicant, these methces are evaluated by the staff for acceptaoility. For new generic methods,~ L
- ne reviewer requests an evaluation by Cp3.
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The values of the parameters used in the analytical model are to be suitably conservative.
The use of the following values is consicered acceptable:
a.
The reactor is initially at 102% of licensed core thermal power (2t allowance for power measurement uncertainty).
b.
Conservative scram characteristics are assumed, i.e., maximum time delay with the most reactive rod held out of the core.
J The core burnup is selected to yield the most limiting combination of moderator c.
temperature coefficient, void coefficient, Coppler coefficient, axial power profile, and radial power distribution.
III. REV!EW PROCEOURES The procedures below are used during both the constmction pemit (CP) and operating license (CL) reviews. During the CP review, the values of system parameters and setpoints used in the,nain.a will be preliminary in nature and subject to change. At the OL review, final values should be used in the analysis, ano the reviewer should comoare these to the limiting safety system settings included in the proposed technical specifications.
The applicant's description of the inadvertent safety or relief valve cpening transient is reviewed by R$8 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 nomally coerating plant instrtenentatien and controls are assumed to function.
2.
The extent to which plant and reactor protection systems are required to function.
3 The credit taken for the functioning of nonnally operating plant systems.
4 The cperation of engineered safety systems that is required.
5.
The extent to which operator actions are required.
If the SAR states that the inadver*ent safety er relief valve opening transient is not as limiting as some other similar transient, the reviewer evaluates the justification presented by the acolicant. If a quantitative analysis of the transient is prest i+.ed in the SAR, the RSS reviewer, with the aid of the EICSB reviewer, reviews the timing of the initiation of those protection, engineered safety, and other systems needed to limit the consequences of the transient to acceptable levels. The RSB reviewer compares the predicted variation of system parameters with various trip and system initiaticn setpoints. Be EICSB reviewer evaluates automatic initiation, actuation delays, possible bypass modes, interlocks, and the feasibility of manual oceration if the SAR states that operator action is needed or excected.
1909 346 15.5.1-3 11/2a/75 um
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To the extent deemed necessary, the RS8 reviewer evaluates the effects of single active failures of systems and components which may alter the course of the transient. In this
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pnase of the review.the system reviews are perferned as described in the star.dard review plans for Chapters 5, 6, 7 and 8 of the SAR. The reviewer considers possible single 9
failures in. systems that replenish or maintain the reactor coolant inventory.
4..
The mathematical models 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
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detemine if these models have been previcusly miewed and fcund acceptable by the staff.
- ~M If not, CPS is requested to initiate a generic review of the model preposed by the applicant.
The values of system parameters and initial core and system conditions used as input to the model are reviewed by RSB. Of particular importance are the reactivity ccefficients and centrol rod worths used by the applicant in his analysis, and the variaticn of moderator temperature, void, and Doppler coefficients of reactivity with core life. The justification provided by the applicant to show that he has selected the core burnuo that yields the minimum margins is evaluated. CPB 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 regarding the maximum pressure in the reactor coolant and main steam systems.
The variation with time during the transient of the core and barrier performance parameters listed in the Event Evaluation Section of Chapter 15 of the Standard : mat (Ref. 6) are reviewed. Values of the more i' 3ortant of these parameters for the tr'isient caused by inadvertent safety or relief valve opening are compared to those predicted for other similarl plants to confirm that they are within the expected range.
IV.
EVAWATION FIN 0!NGS The reviewer verifies that the SAR contains sufficient information and his review supports the following kinds of statements and conclusions, which should be included in the staff's safety evaluation report (SER):
"A number of plant transients can result in a decrease of reactor coolant inventory.
Those that might be expected to occur with moderate frequency are safety or relief valve openings, minor primary pipe breaks, and (in 3WR's) loss of feedwater.* All of 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 systems was the transient. This transient was evaluated by the applicant using a mathematical model that had been previously reviewed and found acceptable by the staff. The parameters used as input to this 'nodel were reviewed and found to be suitably conservative. The results of the analysis of the transient showed that cladding integrity was maintained by ensuring that the minimum departure from
- !ne SER craf t sneuld present ene state-ent for all similar transients.
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nucleate boiling ratio (CNBR)* did not decrease below and that the Caximum pressure within the reactor coolant and main steam systems did not exceed 110% of the design pressures.
"The staff concludes that the plant design is acceptable in regard to transients th '
O are expected to occur with..ioderate frscuency and result in a decreased prirary ecolant inventory."
2.
REFERENCES q
1.
ASME Soiler and Pressure Vessel Code, Section !!I, " Nuclear Power Plant Comp ants,"
Article NB-7000. " Protection Against Overpressure," American Society of Mechanical 5
Engineers.
2.
" Standard Safety Analysis Report - BWR/5," General Electric Ccmpany, April 1973 (under review).
3.
" Reference Safety Analysis Report - RESAR-3," Westinghouse Nuclear Energy Systems, Novemer 1973; and " Reference Safety Analysis Report - RESAR-41" (under review).
4
' System 80 Standard Safety Analysis Report (CESSAR)," Comeustion Engineering, Inc.,
August 1973 (ugder review).
5.
" Standard Nuclear Steam System, 3-SAR-241," Babcock and Wilcox Comoany, FeDruary 1974 (under review).
6.
Regulatory Guide 1.70, " Standard Fermat and Content of Safety Analysis Reports for Nuclear Power Plants," Revision 2.
7.
Standard Review Plan 4.4, " Thermal and Hydraulic Design."
1909 348 15.6.1-5 11/24/75
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1 JAH 16 93 Date:
Serial No.:
IE:20I :TP-79-01 55 2:
TRANSFER OF LEAD RESPCNSIBILITY
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B. K. Grimes, Assistant Director for Engineering and Projects, Divisien of Operating Reactors, Office of Nuclear Reactor Regulation
SUBJECT:
DAVIS BESSE STEAM GENERATOR LEVEL CONTROL DURING AUXILIARY ga FEE 0 WATER OPERATION
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55 RESPONSIBLE ASSISTANT DIRECTOR:
E. L. Jordan, Assistant Director for D
Tecnnical Programs, Division of Reactor Operatins Inspection, Ed Office of Inspection and Enforcement gg na INTRODUCTION:
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==w This confirms and for nalizes a Transfer of Lead Responsibility which was
$M verbally made December 23, 1978. Timely involve ent and assistance by.
$1 the Office of Nuclear Reactor Regulation (NRR) on this issue as it unfolded gjg is acknowledged and appreciated.
R3 DESCRIPTION OF ITEM REQUIRING RESOLUTION:
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l During.startup testing of the Davis Besse reactor the pressurizer level
@q fell below the level indicating range, following a loss of effsite power t
and scram from 3% power.
The level drop was the result of primary coolant y[
contraction caused by the large volume of cold auxiliary feedwater that is
!=23 pumped in to bring the steam generator levels up to 120 inches frcm the
.[555 normal operating level of about 30 inches at this power.
Office of Inspection CM and Enforcement (IE) review of the licensee's planned ccrrective actions E2 identified potential unreviewed safety questions sir.ca the small break analyses were performed assuming a 120 inch steam generation level and the p6 5-actual pressurizer level drop for a loss of offsite power at 100'.' power if was not determined.
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E5 c5E9 The licensee intends to permanently correct the pressurizer level problem it during a future.,hutdown period by reducing the automatic control level of gg the auxiliary feedwater system from 120 inches to 35 inches unless a safecy
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irijection is received.
Until this change can be implemented, in the event Z
of loss of offsite power, the steam generator levels are to be held at 35 inches by manual centrol, in accordanca with a revised procedure.
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, JAN 1619R procedure calls for maintaining the steam generator control at 120 inches f.
in the event of a safety injection signal combined with a loss of offsite power.
The li.censee reports that a recent natural circulation test at Davis Besse demonstrated that the 35 inch level provides adequate natural circulation for decay heat removal (en. closed).
The licensee believes that the temporary procedure change and the planned 5l pemanent fix do not involve an unreviewed safety question and can, there-
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fore, be made without NRC approval, in accordarce with 10CFR 50.59.
- However, 5::t:;
several questions wer; raised during joint discussions betweer :i, NRR
. a-and the licensee on December 23, 1978.
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RECOMMENDATIONS AND PROPOSED COURSE OF ACTION:
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1.
NRR will detemine the acceptability of the temporary use of operator EC action to control steam generator water level at 35 inches in the event as of a loss of offsite power and to control at 120 inches in the event
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of a safety injection following a loss of offsite power.
2.
NRR will detemine the acceptability of the licensee's pemanent corrective f5 action.
3.
IE will verify adequacy of administrative controls temporarily imposed E5 by the licensee in connection with item 1.
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4.
IE will verify compliance with any requirements established by NRR ei-and examine any plant modifications associated with a pemanent fix m.
to this problem.
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STAFF CONTACT:
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IE:
D. C. Kirkpatrick (28180)
Ec NRR:
G. 5. Vissinc (27435)
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CONCURRENCE:
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G; M_./dcrcan, Assistant Director for Tecnnical Programs,
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Division'
- Reactor Operations Inspection, IE
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Srie:f %. Grimes, Assistant 01 rector for Engineering anc Projects, Date
=e Divisien of Operating Reactors, NRR 5=!
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3-JAN 16 573 G
Enclosure:
Ltr, Toledo Edison to R. W. Reid d::i 12/22/78
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cc:
R. S. Boyd, DPM
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V. Stellr, DDR a
D. G. E1Ienhut, DOR
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R. W. Reid, DOR C
G. '!issing, DOR J. G. Davis, IE s;a N. C. Mo.;eley, IE Mi H. D. Thornburg, IE
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G. R. Klingler, IE 555 G. W. Reinr::uth, IE
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B. H. Grier, RI
""h J. P. O'Reilly, RII J. G. Keppler, RIII
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K. V. Seyfri*., RIV Ej R. H. Engelken, RV g.?
K. Terney, MPA Las L. V. Gossick, EDO Wii R. J. Mattson, DSS El L. Shao, RES
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