ML20040A138
| ML20040A138 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/12/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TASK-15-09, TASK-15-9, TASK-RR LSO5-82-01-031, LSO5-82-1-31, NUDOCS 8201200396 | |
| Download: ML20040A138 (6) | |
Text
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January 12, 1982 Docket No. 50-245 co LS05 01-031
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Mr. W. G. Counsil, Vice President 2
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Dear Mr. Counsil:
SUBJECT:
MILLSTONE 1 - SEP TOPIC XV-9, STARTUP OF AN INACTIVE LOOP OR RECIRCULATION LOOP AT AN INCORRECT TEWERATURE AND FLOW CONTROLLER MALFUNCTION CAUSING AN INCREASE IN BWR CORE FLOW RATE By letter dated June 30,1981, you submitted a safety assessment mport for the above topic. The staff has reviewed this assessment and our conclusions are presented in the enclosed safety evaluation report, which conpletes the review of this topic for Millstone 1.
This evaluation will be a basic input to the integrated assessment for your facility. The evaluation may be mvised in the future if your facility design is changed or if NRC criteria relating to this tppic are modified before the integrated assessment is completed.
.s G,o W Sincerely, Orir:inal signed byr bitt $56 Dennis M. Crutchfield, Chief Operating Reactors Branch No. S M *.
Division of Licensing D. IEU "!'
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As stated cc w/ enclosure:
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SYSTE?tATfC EVALUATI0'l PROGRA?1 TOP!C XV-9 STARTUP OF AN INACTIVE LOOP OR RECIRCULATION LOOP AT AN INCORRECT TEMPERATURE AND FLOW CONTROLLER MALFUNCTION CAUSING AN INCREASE IN BWR CORE FLOW RATE MILLSTONE 1 I.
INTRODUCTION The objective of th'is review is to assure that the consequences of core flow in-crease transients are acceptable, i.e., that the increase in core flow or the in-troduction of cooler or 'deborated water into the core does not lead to an unaccept-able loss of fuel clad integrity or overpressurization of.the primary system.
The improper startup of an idle recirculation loop would cause a power increase due to a combination of a negative moderator c'oefficient addition and increasing recir-culation flow.
A flow controller malfunction can cause the scoop tube position to move at its maximum speed in the direction of increasing pump flow.
This increasing pump flow will sweep voids at a faster rate and therefore cause an increase in neu-tron flux.
The review consists of evaluating the licensee's analysis of the sequence of events, the analytical model, the values of parameters used in the analytical model and the predicted consequences of the transients.
II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each' applicant for a construction per-mit or operating license provide an analysis and -evaluation 'of the design and per-formance of structures, systems', and components of the facility with the objective of assessing the risk to public health and safety resulting from operat, ion of the facility, including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility.
Section 50.36 of 10 CFR Part 50 requires the Technical Specifications to include d
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against the uncontrolled release of radioactivity.
The General Design Criterion (Appendix A to 10 CFR Part 50) set forth tne criter;r for the design of water-cooled reactors.
GDC 10 " Reactor Design" requires that the core and associated cooling, control and protection systems.,be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during normal operation, including the effects of anticipated operational occurrences.
Gbc 15 " Reactor Coolant System Design" requires that the reactor coolant and asso-ciated protection systems be designed with sufficient margin to assure that the design conditions of t.he reactor coolant pressure boundary are not exceeded during normal operation, including the ' effects of anticipated operational occurrences.
j GDC 20 " Protection System Functions" requires that the protection system be designed to initiate automatically.the operation of reactivity control syptems to assure.
that specified acceptable fuel design limits are not exceeded as a result of anti-cipated operational occurrences.
GDC 26 " Reactivity Control Sy' stem Redundance and Capability" requires that the reactivity control' system be capable of reliably controlling reactivity changes to assure that under conditions of nonnal operation, including anticipated operational occurrences, and with appropriate mar, gin for malfunctions such as stuck rods, spe-cified acceptable fuel design limits are not exceeded.
GDC 28 " Reactivity Limits" requires that the reactivity control systems be designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that.the effects of postulated reactivity acci. dents can neither (1) result
,in damage to the reactor coolant pressure boundary greater than limited local yield-ing nor (2) sufficiently disturb the core, its support structures or other reactor
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i pressure vessel internals to impair significantly the capability to cool the core.
III.
R_ ELATED _ SAFETY TOPICS l
Various other SEP topics evaluate such items as the reactor p'rotection system.
The effects of single failures on safe shutdown capability are considered under Topic
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VII-3.
4 VI.
REVIElf GUIDELINES The review is conducted in accordance with SRP 15.4.4 and 15.4.5.
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.The evaluation includes review of the analysis for the event and identification of s
l the features in the plant that mitigate the consequences of the event as well as the ability of these systems to function as required.
The' extent to which operator action is required is also evaluated.
Deviations from the criteria specified in the Standard Review Plan are identified.
V.
EVALUATION i
Both of the subject transients were assessed in the General Electric generic reload topical (Reference 1) and it was determined that the inadvertent flow increase tran-sient was the most limiting of the transients at reduced flow.
For the FSAR analysis of the inactive loop startup, the inactive loop is assumed' to 0
be filled with 100 F water.
The analysis assumes 70% of full power and 40% core
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flow, with 10% flow reversed through' the idle loop.
The analyses were performed at the end of equilibrium fuel. cycle exposure condition in which the scram and void characteristics would be the worst.
The minimum critical heat flux ratio (MCHFR)
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is 1,25, which is above the limit of 1.0.
The general analytical methods have been approved by the staff (Reference 2),
For the flos controller malfunction transient, the assumed initial conditions are 4
65% power and 50% flow, The failed motor 9,enerator set speed controller causes 4
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F-P the scoop tube position to move at a rate of 10%/second for nine seconds.
The use factors to adjust the MCPR operating limits will ensure that tne 0 -
of generic Kf factors erating limit MCPR will not be violated for this event.
Appropriate Kf have been reviewed and approved in. Reference 2.
Additional discussion of these events was provided in Reference 3 in which appro-priate operator actions were identified.
VI.
C0tiCLUS10 tis As part of the SEP review of Millstone 1, the startup of an inactive 1 cop and flow controller malfunction transient were reviewed against the acceptance criteria of SRP Section 15.4.'4 and 15.4.5.
The initial conditions and analytical methods have been reviewed and.found to conform to the requirements of the SRP.
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E REFERENCES 1.
NEDE-240ll-P " Generic Reload Fuel Application, with Revisions" May,1977 -
March, 1978.
2.
Staff Evaluation GE Generic Reload Application, May 12, 1978.
s 3.
NED0-24078-A " Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors, Volume 1, August,1979, Revision 1 - December,1980.
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