ML20038C206
| ML20038C206 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 12/04/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TASK-15-07, TASK-15-7, TASK-RR LSO5-81-12-009, LSO5-81-12-9, NUDOCS 8112100264 | |
| Download: ML20038C206 (9) | |
Text
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r Deces.ber 4,1981 N
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Docket No. 50-245 N
LS05 12-009
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7 Mr. W. G. Counsil. Vice President
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k Huclear Engineering and Operations S.,
AU Northeast Nuclear Energy Company
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Post Office Box 270 4//T_t. 4.
Hartford, Connecticut 06101
Dear Mr. Counsil:
SUBJECT:
MILLSTONE 1 - SEP TOPIC XV-7, REACTOR COOLANT PUMP ROTOR SEIZURE AND REACTOR COOLANT PUMP SHAFT BREAK By letter dated June 30, 1981, you submitted a safety assessment report for the above topic. The staff has reviewed this assessment and our conclusions are presented in the enclosed safety evaluation report, which completes this topic for Millstone 1.
This evaluation will be a basic input to the integrated assessment for your facility. The evaluation may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.
Sincerely, Dennis H. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing
Enclosure:
As stated cc w/ enclosure:
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MILLSTONE 1 Docket No. 50-245
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Mr. W. G. Counsil CC William H. Cuddy, Esqr Connecticut Energy Agency Day, Berry & Howard ATTN: " Assistant Director Counselors at Law Research and Policy One Constitution Plaza Development Hartford, Connecticut 06103 Department of Planning and Energy Policy Natural Resources Defense Council 20 Grand Street 91715th Street, N. W.
Hartford, Connecticut 06106 Washington, D. C.
20005 Northeast Nuclear Energy Company ATTN: Superintendent Mi.11 stone Plant i
P. O. Box 128 Waterford, Connecticut 06385 Mr. Richard T. Laudenat Manager, Generation Facilities Licensing Northeast Utilities Service Company.
P. O. Box 270 l
Hartford, Connecticut 06101 Resident Inspector c/o U. S. NRC P. O. Box Drawer KK i
l Niantic, Connecticut 06357 l
l Waterford Public Library Rope Ferry Road, Route 156' Waterford, Connecticut 06385 First Selectman of the Town i
of Waterford l
Hall of Records l
200 Boston Post Road Waterford, Connecticut 06385 John F. O' eka p
Systems Superintendent Northeast Utilities Service Company P. O. Box 270 l
Hartford, Connecticut 06101 U. S. Environmental Protection Agency Region I Office ATTN: EIS C0ORDINATOR JFK Federal Building Boston, Massachusetts 02203
MILLSTONE 1 SEP TOPIC XV-7i LOSS ~~0F FORCED RECIRCULATION FLOW, PUMP ROTOR SEIZURE ~ "
AND. PUMP S.H. AFT B_REAK,,, _,,, _ __
Loss of Forced Reactor Coolant Flow Includino ' Trip of Pump Motor and Flow Controller Malfunctions I.
110RODUCTION A decrease in'ieactor ccolcat flow occuring while the plcnt is at power could result in a degradation of core heat transfer. A r,csulting increase i
in. fuel temperature and ccccmpanying fuel damage could then result if speci-l ficd acceptable fuel damage limits are exceeded during the transient. A l
i number of transients that are expected to occur with moderate frequency and i
that result in a decrease in forced reactor coolant flow rat'e are addressed -:
For boiling water reactors (BWRs), partial and
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complete recirculation pump trips and malfunctions of the recirculation flow controller to cause decreasing flow are reviewed.
II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for an operating license provide an analysis and evaluatien of the design and performance of structures, systems and components of the facility with the objective of as-sessing the risk to public iiealth an'd safety re'sulting frem operation of the facility.
The loss of forced reactor coolant flow is one of the postulated transients used to evaluate the adequacy of these structures, systems and components with respect to the public health and safety.
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The General Design Criteria (Appendix A to 10 CFR Part 50) establish' minimum rce,uirements for the principal. design criteria for water-cooled reactors.
The, staff accepttnce criteria are based on meeting the relevant requirements of the following regulations:
A.
Ccneral Design Criterion 10 (Ref.1), as it relates to the reactor coolant system being designed with appropriate margin to assure that specirkd acci.ptcble fuel design limits are not excccded during non..al ope ations including cnticipated operational occurrences.,
.B.
Cencral' Design Criterion 15 (Ref. 2), as it relates to the reactor coolant s'ystem and its associated auxiliaries being designed with appropriate margin to assure that the pressure boundary will not be brceched during normal operations including anticipated opera-tional occurrcnces.
C.
General Design Criterion 26 (Ref. 3) as it relates to the reliable control of reactivity changes to assure that specified acceptable fuel design limits are not exceeded, including antic'ipated opera-tional occurrences. This is accomplished by assuring that appro-priate margin for malfunctions, such as stuck rods, are accounted for.
The specific critoria necessary to meet the relevant requirements of GDC 10, 15 and 26 for incidents of moderate frequency are:
a.
Pressure in the reactor coolant and main steam systems should be maintained below 110% of the design values.
b.
Fuel cladding integrity shall be maintain.ed by ensuring that the minimum DNBR remains above the 95/95 DNBR limit for PWRs and the CPR remains above the MCPR safety limit for BURS based on accept-able correlations (see SRP Section 4.4).
An incident of moderate fecquency should not generate a more serious c.
plant condit. ion wii.hout other faults occurring independently.
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d.
An incident of iroddrate frequoncy in combination with any single ac-tiv'e component failure, or single operator' error, shall be considered and is an event for which an cstimate. of~ the number of potential fuel failures shall be 'provided for radiol.ogical dose calculations.
For such accidents, the number of fuel
- failures must be assumed for all rods for which the DNBR or CPR falls below those values cited above for cladding integrity unless it can be shown, based on an acceptable fuel damage model (see SRP Section 4.2), that fewer failures occur.
There shall be no loss of function of any fission product barrier other than the fuel cladding.
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III.
RELATED SAFETY TOPICS I
Varicus other SEP topics evaluate such items as the reactor protec-tien system.
The effects of single failures on safe shutdown capabiP:.
l lity are considered under Topic VII-3.
I V.
REVIEW GUIDEl.It!ES
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The revicw is condected in cccordance with SRP sections 15.3'.1 and 15.3 The evaluation includes reviens of_the _ analysis _for the_ event _and identi_-
fication of the features in the plant that mitigate the consequcnces of the event as well as the ability of these systems to function as required. The extent to which operator action is require'd is also evaluated.
Deviations from the criteria specified in the Standard t
Review Plan are identified.
V.
EVALUATIOR The licensee in their letter of June 30, 1981 provided the re~sults of an analysis for the subject topic.
The analysis indicates that a loss of reactor coolant flow can result from loss of power to the pump, failure of drive motor connections, M-G set breakers or pump failure.
The decreasing core flow causes a core heat-up due to the flow-power mismatch.
The increased void formation inserts negative reactivity to drop power back to a level compatible with.the lower flow. No reactor trips occur due to the decreased flow. The rotating inertia of the recirculation drive equipmeht slows the flow coastdown so that equil.ibrium conditions are reestablished without exceeding fuel limits.
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The licensee has not provided the results of an aha. lysis for this event
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in. combination with a single failure. Since this event does not cause a reactor trip or any engineered safety feature initiation during this transient, we could not identify any single failure which would lead to unacceptable results.
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CONCLUSIONS The staff concludes that the Millstone 1 plant design with regard to transients that are expected to occur during plant life and result in a loss of decrease in forced reactor coolant flow is acceptable and meets the relevant requirements of General Design Criteria 10,15 and
- 26. This conclusion is based on the following:
1.
The applicant has met the requirements of GDC 10 and 26 with respect to demonstrating that the specified acceptable fuel design limits are not exceeded for this event. This requirement has been met since the results of the analysis shcucd that the thermal margin limits (MCHFR) are satisfied.
l 2.
Y;.e 0; plic.:nt has c.ct t! e requitw.zats of 00C.15 with rysrect to l
<!::;.:ua,tcal.i.;g that the reactor coolant p;usure bounkry limits ha /c not b ca r;tecedcd for i.his.ncut.
This rcquireirent has been j
nt since ihe r.nalysis showed that L:.e eaxir."a pressure of the
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t. c.ciuc coolant ad taia stca.a syi.os did not exccad 110% of the i
<!csiga ;.rctsure.
l 3.
The applicant has ict the requir =1ts of GDC 26 with respect to the ci pcbility of U,e reactivity caatrel systra to provide adcquate ceatcol of reactivity during ihis event !iile %cludirg appropriate v.argin for stuck rods since the specific acceptabic fuel design limits were not exceeded.
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1 Recirculation Pump Seizure I.
Introduction.
During the pump seizure event, reactor coolant flow drops rapidly and core power
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...... a-decreases due to additional void formation. The reactor trip level is not reached during this event.
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The licensee has performed an analysis of the recirculation pump seizure event (Ref.1).
The analysis assumes the reactor at 100% power.
II.
Review Criteria Section 50.34 of 10 CFR Part 50 requires that each. applicant for. an operating license provide an analysis and evaluation of the design and performance of structures, systems and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility.
The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum i
requirements for the principal design criteria for water-cooled reactors.
GDC 27 " Combined Reactivity Control System Capability," requires that the i
reactivity control systems, in conjunction with poison addition by the emergency core cooling system, has the capability to reliably contral I
reactivity changes to assure that under postulated accident conditions, and i
with appropriate margin for stuck rods the capability to cool the core is maintained.
1 h
GDC 28 " Reactivity imits" requires that the reactivity control systems be designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that the eff,ects of postulated.reacti.vity accidents can neither (1) result in damage to. the reactor coolant pressure boundary ' reater t'han limited local yielding nor (2)sufficiently disturb the g
core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core.
GDC 31 " Fracture Prevention of Reactor Coolant Pressure Bou'ndary" requires d
that the boundary be designed with sufficient margin to assure that when stressed under operating, maintenance, testing and postulated accident condi-tions (1) the boundary behaves in a nonbrittle manner and (2),the probability of rapdily propagating fractures is minimized.
III. Related Safety Topics Various other SEP topics evaluate such items as the reactor protection system.
The effects df single failure on safe shutdown capability are considered under Topic VII-3.
IV.
Review Guidelines The review is conducted in accordance with SRP l'5.3.1,15.3.2,15.3.3,15.3.4.
The evaluation includes review of the analysis.for the event and identifica-tion of the features in the plant that mitigate the consequences of the ev.en.t,3, as well a,s th,e ab,iIity of these systems to function as required. The extent, to which operator act' ion is required is also evaluated.
Deviations from the criteria specified in the Standara Review Plan are identified.
V.
Evaluation The r,esu3ts of the licensee's analysis indicate that the reactor does ngt trip during this event.
The lowest MCHFR reached is 1.13 and the pressure ih slightly reduced from normal operating pressure. Therefore, the additional over pressure that may result from an initial reactor power of 102%, required by the SRP section 15.3.3, as opposed to the 100% assumed in the analysis would not be signifidant.
A limiting single failure and loss of offsite power are ndt identified for
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this analysis.
However, since the reactor is not tripped during this event, a sin'gle failure cannot be identified that may alter the course of mitigation.
Additionally, loss of offsite powhr wod1d further reduce the s.everity of the consequences.
VI.
Conclusion Therefore, we conclude that the analysis of the rotor seizure event is accepta bl e.
Recirculation Pump Shaft Break l
j The recirculation pump shaft-break transient has not been analyzed by the licensee, because its consequences are less severe than the consequences during the pump s,eizure event. Therefore, separate,, analysis is not re.
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quired for this event.
References 1.
Letter fgom W. G. Counsil and R. P. Werner, Northeast Utilities, to Director of Ndclear Reactor Regulation, Attn:
D. M. Crutchfield, NRC,
Subject:
J Millstone Nuclear Power Station, Unit No.1, SEP Section XV Topic 8; Design Basis Events, dated June.30, 1981.
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