Text

Forwards Safety Evaluation of Util 810630 Safety Assessment Re SEP Topic XV-1, Decrease in Feedwater Temp,Increase in Feedwater Flow & Increase in Steam Flow. Tech Specs Re Inoperability of Bypass Sys Required
	ML20039E535
Person / Time
Site:	Millstone
Issue date:	12/31/1981
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	Counsil W; NORTHEAST NUCLEAR ENERGY CO.
References
	TASK-15-01, TASK-15-1, TASK-RR LSO5-81-12-105, NUDOCS 8201110003
	Download: ML20039E535 (15)
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Q Q-bpo ld December 31, 1981 Docket No. 50-245 m

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<'s REcE;ygg Mr. W. G. Counsil, Vice President

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!!uclear Engineering and Operations Northeast Nuclear Energy Company sacav Post Office Box 270 c>

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Dear Mr. Counsil:

SUBJECT:

MILLSTONE 1 - SEP TOPIC XV-1 DECREASE IN FEEDWATER TEMPERATURE, INCREASE Ill FEEDWATER FLOW AND INCREASE Ifl STEAM FLOW In your letter dated June 30, 1981, you submitted a safety assessment report on the above topic. The staff has reviewed your assessment and our conclusions are presented in the enclosed safety evaluation report.

Our report completes this topic evaluation for 11111 stone 1.

5 As noted in the evaluation of the feedwater controller malfunction event, the staff will require Technical Specifications changes to conform with current licensing practice if credit is to be given for operation of the turbine bypass system in the analyses.

The enclosed safety evaluation will be a basic input to the integrated safety assessment for your facility. The asses e nt may be revised in the future if your facility design is changed.1 i f i:RC criteria relating to this topic are modified before the inteo9W tssesment is completed.

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Dennis M. Crutchfield, Chief l

Operating Reactors Branch Na. 5 Division of Licensing

Enclosure:

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William H. Cuddy, Esquire 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, Connet.ticut 06106 Washington, D. C.

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Northeast Nuclear Energy Company ATTN: Superintendent Millstone Plant i

P. O. Box 128 l

Waterford, Connecticut 06385 Mr. Richard T. Laudenat Manager, Generation Facilities Licensing l

Northeast Utilitics Service Company.

P. O. Box 270 06101 Hartford, Connecticut Resident Inspector i

c/o U. S. NRC-P. O. Box Draw 6r KK j

Niantic, Connecticut, 06357 Waterford Public Library Rope Ferry Road, Route 156 4

Waterford, Connecticut 06385 4

i First Selectman of the Town

" of Waterford Hall of Records 200 Boston Post Road Waterford, Connecticut 06385 John F. Opeka Systems Superintendent 1

Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 U. S. Environmental Protection Agency Region 1 Office ATTN: EIS COORDINATOR JFK Federal Building Boston, Massachusetts 02203 e

SYSTEMATIC EVALUATION PROGRAM TOPIC XV-1 l

MILLSTONE 1 l

I TOPICcXV-It- - DECREASE IN FEEDWATER TEMPERATURE, INCREASE IN FEEDWATER l

FLOW, INCREASE IN STEAM FLOW 4

DECREASE IN FEEDWATER TEMPERATURE W-I.

INTRODUCTION Feedwater heating can be lost by closure of the steam extraction lines to the heaters or the bypassing of feedwater around the heaters.

In either case the reactor vessel receives cocler feedwater and there is an increase in core inlet subcooling.

The decrease in coolant void fraction and the negative void react-ivity coefficient result in a gradual initial incru e in reactor power.

The event could occur with the reactor in either manual or automatic control mode.

In the automati,c control mode, there is some compensation for reactor power increase by modul'ation of core flow and the event generally is less severe.

In the manual control mode, and assuming no corrective operator actions, the reactor rower could either reach a higher equilibrium value below the scram setpoint or increase sufficiently to cause automatic scram on high neutron flux.

The power history depends on both the assumed maximum decrease in feedwater temperature and the feedwater temperature time constant.

The loss of feedwater heating event results in a mild transient in which the fuel surface heat flux increases to a maximum value below that corresponding to steady state operation at the scram setpoint.

This increase in power to the coolant is partially offset by the beneficial effect of the increased core inlet subcooling on the critical power ratio.

However, this event can be one of the limiting events l

with respect to minimum critical power ratio, and is considered in reload analyses.

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II. REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objec-tive of assessing the risk to public health and safety resulting from operation of the facility, including detemination 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 safety limits which protect the integrity of the physical barriers which guard against the uncontrolled release of radioactivity.

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The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum re-quirements for the principal design criteria for water-ccoled reactors.

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GDC 10 " Reactor Design" requires that the core and associated coolant, control and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during normal operacion, including the effects of anticipated operational occurence.

GDC 15 " Reactor Coolant System Design" requires that the reactor coolant and associated protection systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation, including the effects of anticipated operational occurrences.

I GDC 26 " Reactivity Control System Redundance and Capability" requires that the l

mactivity control systems be capable of reliably controlling reactivity changes to assum that under conditions of normal operation, including anticipated l

4 operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded.

i III. RELATED SAFETY TOPICS Various other SEP topics evaluate such items as the reactor protection system.

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i The effects of single failures on safe shutdown capability are considered under i

5 Topic VII-3.

IV. REVIEW GUIDELINES 4

The review is conducted in accordance with SRP 15.1.1,15.1.2,15.1.3 and 15.1.4.

The evaluation includes review of the analysis for the event and identification of the features in the plant that mitigate the consequences t f 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 In reference 1, the licensee reported on calculations of reactor response to a loss of feedwater heating event in which feedwater temperature decreased by 100 F.

Reactor power increased to a new equilibrium value of about 118 percent with a corresponding decrease in critical power ratio of 0.15.

This event was analyzed for an initial power of 100 percent instead of 102 percent as required by SRP Section 15.1.1.

Use of the higher initial power could result in a slightly larger decrease in critical power ratio. However, the limiting transient for Millstone Unit 1 is generator load rejection without bypass which results in a decrease in critical power ratio of about 0.30.

Hence reevaluation of the loss of feedwater heating event with 102 percent initial power would not change the operating minimum critical power ratio limit.

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VI. CONCLUSIONS h

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As part of the SEP review for Millstone 1, we have evaluated the licensee's j

analysis of the loss of feedwater heating event. We conclude that the loss of l

feedwater heating event is bounded by generator load rejection without bypass.

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We, therefore, find. results acceptable even though an initial power of 100%

was assumed instead of 102% as required by the SRP acceptance criteria.

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l REFERENCE 1.

Y1003J01 A09, " Supplemental Reload Licensing Submittal for Millstone Unit 1 Reload 7," June 1980.

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i MILLSTONE 1, SEP TOPIC XV-1 EVALUATION INCREASE IN FEEDWATER FLOW I.

INTRODUCTION Failure of the feedwater controller to maximum demand results in an increase in re-actor power and vessel inventory.

There is a gradual initial increase in power be-cause of the increased core inlet subcooling and the negative void coefficient of reactivity.

The steam /feedwater flow mismatch leads to a high vessel water level trip of the main turbine.

The turbine trip results in a pressurization transient, with attendant power transient, which is mitigated by reactor scram due to turbine stop/ control valve closure and initiation of the turbine bypass system.

The limi-ting conditions occur during the pressurization portion of the overall event.

A feedwater controller failure at partial power gives a larger steam /feedwater flow i

mismatch.

However, failure at rated power can be more severe in terms of m'aximum reacter pressure and minimum critical power ratio. A feedwater control failure event at rated power is similar to the turbine trip e, vent at rated power with tur-bine bypass operable.

However, for the feedwater controller event, the turbine trip signal occurs when the reactor is at above rated power.

Hence this event can be

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limitin.g with respect to minimum critical power and is evaluated in reload analyses.

i To meet current licensing criteria, surveillance of the turbine bypass system is required. Since the bypass system was assumed to operate in the analysis of this event, limitations to either reactor power or minimum critical power ratio would be required in the Technical Specifications to cover the case where the bypass system i

i is found inoperable.

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II. REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systens, and components of the facility with the object-ive of assessing the risk to public health and safety resulting from operation of the facility, including determination of the margins of safety during normal oper-ations and transients conditions anticipated during the life of the facility.

Section 50.36 of 10 CFR Part 50 requires the Technical Specifications to include safety limits which protect the integrity of the physical barriers which guard j

against the uncontrolled release of radioactivity.

The General Oasign Criteria (Appendix A to 10 CFR Part 50) establish minimum re-quirements for the principal design criteria for water-cooled reactors.

i GDC 10 " Reactor Design" requires that the core and associated coolant, control i

and protection systens be designed with appropriate margin to assure that speci-fied acceptable fuel design limits are not exceeded during normal operation inclu-t ding the effects of anticipated operational occurence.

GDC 15 " Reactor Coolant System Design" requires that the reactor coolant and associated protection systems be assigned with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation, including the effects of anticipated operational occurrences.

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3 GDC 26 " Reactivity Control System Redundance and Capability" require; that the reactivity co'ntroi systems be capable of reliably controlling reactivity changes j

to assure that under conditions of normal operation, including anticipated operatio'nal occurrences, and with appropriate margin for malfunctions such as.

stuck rods, specified acceptable fuel design limits are not exceeded.

III.

_REl.ATED SAFETY TOPICS l

Various other SEP topics evaluate such items as the reactor pr[)tection system.

The effects of single failures on safe..s'hutdown capability are considered under Topic VII-3.

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IV.

REVIEW GUIDELINES 4

The review is conducted in accordance with SRR 15.1.1, 15.1.2, 15.1.3 and

,15.1. 4.

The evaluation includes review of the a'nalysis for the event and identification of the features in the plant that mitigate the consequences of the event as

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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.

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V. EVALUATION In Reference 1, the licensee evaluated the consequences of a feedwater controller failure leading to a:i increase in feedwater flow to 110%. The~ initial power was 100% instead of 102% as required in SRP Section 15.1.2.

However, the reduction in critical power ratio was only 0.09 compared to a reduction of 0.30 for load rejection without bypass. Hence, reevaluation of this event for an initial power of 102% would not result in a reduction in the operating minimum critical power ratio limit.

VI. CONCLUSIONS-As part of the SEP review of Millstone 1, we have evaluated the licensee's analysis of a feedWater controller failure. We conclude that this event is bounded by load rejection without bypass. We, therefore, find the results acceptable even though an initial power of 100% was assumed instead of 102%

as required by the SRP. acceptance criteria.

To meet current licensing criteria, surveillance of the turbine bypass system is required. Since the bypass system was assumed to operate in the analysis of this event, limitations to either reactor power or minimum critical power ratio would be required in the Technical Specifications to cover the case when the bypass system is found inoperable.

REFERENCES 1.

Y1003J01A09, " Supplemental Reload Licensing Submittal for Millstone Unit 1. Reload 7", June 1980.

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i MILLSTONE 1, SEP TOPIC XV-1 EVALUATION INCREASE IN STEAM FLOW I.

INTRODUCTION j

Failure of the pressure regulator in an open position results in full opening of the turbine admission valves and partial opening of the turbine bypass valver.

The total steam flow rate resulting from the regulator failure is limited to approximately 110 percent of rated flow by a maximum flow limiter. The increased steam flow rate results in a drop in reactor pressure and inventory.

The increase -

in core void fraction produces an initial decrease in core power and increase in vessel level.

The vessel level increase can cause trip of the main turbine.

Re-actor scram then results from turbine stop/ control valves closure.

If the turbine trip signal or high water level s not reached, an MSIV closure on low steam pressure occurs.

Reactor scram then results from position switches on the MSIV's.

Since the turbine trip or MSIV closure occurs at relatively low reactor pressure and power, the pressure regulator failure event is'not of consequence with respect to peak system pressure or minimum critical power ratic.

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REVIEW CRITERIA l

f Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the ' facility with ',

s-the objective of assessing the risk to public., health and safety resulting from '

operation 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 C'FR Part 50 requires the Tgc:hnical Specific'ations to include safety limits which protect the integrity of the physical barriers whic'h guard against the uncoctrolled, release of radioactivity.

The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum reactors.

requirements for the principal design criteria foi-water-cooled s

GDC 10 " Reactor Design" requires that the core and assoc'iated coolant, control and protecticin 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 occurrence.

' CDC 15 " Reactor Coolant System Design" requires that the reactor coolant and associated protection sysfems be designed with sufficient mar. gin to 'as.sure that the design cotiditicos of the reactor coolant pressure boundary are not. exceeded' ~

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during normal opera *> O n," including the effects of antiCihated operutional ~~ ~ \\

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occurrences.

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GDC 26 " Reactivity Control System Redundance and Capability" requires that the reactivity control systems be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operatio'nal occurrences, and with appropriate margin for malfunctions such as stuck ruas, specified acceptable fuel design limits are not exceeded.

4 III.

ML20039E535

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Dear Mr. Counsil:

SUBJECT:

Enclosure:

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ML20039E535

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Dear Mr. Counsil:

SUBJECT:

Enclosure:

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