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Forwards Final Safety Evaluation of SEP Topic XV-1 Re Decrease in Feedwater Temp,Increase in Feedwater Flow & Increase in Steam Flow.Results of Util 820610 Reanalysis of Increase in Feedwater Flow Comply W/Current Criteria
	ML20054K053
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	06/25/1982
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	Vandewalde D; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
References
	TASK-15-01, TASK-15-1, TASK-RR LSO5-82-06-101, LSO5-82-6-101, NUDOCS 8206300347
	Download: ML20054K053 (15)
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i June 25,1982 Docket No. 50-155 L505-82 06-101 s

Mr. David J. VandeWalle Nuclear Licensing Administrator Consumers Power Company 1945 W. Barna11 Road Jackson, Michigan 49201

Dear Mr. VandeWalle:

SUBJECT:

BIG ROCK POINT - SEP TOPIC XV-1, DECREASE IN FEEDWATER TEMPERATURE, INCREASE IN FEEDWATER FLOW AND INCREASE IN STEAM FLOW By letter dated March 4,1982, the staff issued a safetyrevaluation on SEP Topic XV-1 in which we concluded that Technical Specification changes would be required if credit was to be given for operation of the turbine bypass system in the analysis.

Your letter of June 10, 1982, provided a reanalysis of the increase 5g in feedwater flow event without operation of the bypass valve. As-noted in the enclosed safety evaluation of Topic XV-1, we conclude that b54"g,g')

the results of this analysis comply with current licensing criteria, The enclosed final safety evaluation will be a basic input to the inte-grated safety assessment for your facility. The assessment may be revised in the future if your facility design is changed or if NRC $ Db*.

criteria relating:to this topic are modified before the integrated f

I assessment is completed.

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/E Sincerely,

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Dennis M. Crutchfield, Chief jgj6ggCK05000155 a47 820625 Operating Reactors Branch No. 5 Division of Licensing P

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Revised 1/21/82.

Mr. David J. VandeWallo cc Mr. Paul A. Perry, Secretary U. S. Environmental Protection

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Co'nsumers Power Company Agency 21'2 West Michigan Avenue Federal Activities Branch Jackson, Michigan 49201 Re) ion V Office ATTN:

Regional Radiation Representative Judd L. Bacon, Esquire 230 South Dearborn Street Consumers Power Company Chicago, Ill'inois 106b4-

,212 West Michigan Avenue, Jackson, Michigan 49201,

Peter B. Bloch, Chairman Atomic Safety and Licensing Board Joseph Gallo, Esquire U. S. Nuclear Regulatory Commission Isham, Lincoln & Beale 1120 Connecticut Avenue Washington, D. C.

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Room 325 Dr. Oscar H. Earis Washington, D. C.

20036 Ato.nic Safety and Licensing Board _.

U. S3 Nuclear. Regulatory Commission Peter W. Steketee, Esquire Washington, D. C.

20555 505 Peoples Building Grand Rapids, Michigan 49503 Mr. Frederick J. Shon Atomic Safety and Licensing Board

. Alan S. Rosenthal, Esq., Chairman U. S. Nu~ clear Regulatory Commission

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Atomic Safety & Licensing Appeal Board Washington, D. C.

20555 U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Rig Roek-Point Nuclear Power Plant

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Isr-C. J. Hartman Mr. John O'Neill, II

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PTant Superintendent.

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Christa-Maria

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Route 2, Box 108C Charlevoix, Michigan -49720 Charlevoix, Michigan 49720 William J. Scanlon, Esquire 2034 Pauline Boulevard Chairman County Board of Supervisors Ann Arbor, Michigan 48103 Charlevoix County Charlevoix, Michigan 49720 Resident Inspector I'"

Big Rock Point Plant l

' Office 'of the Governor (2) c/o U.S. NRC

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Room i - Capitol Building RR #3, Box 600

__ _ Lansing, Michigan 48913 Charlevoix, Michigan 49720

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Herbert Semmel Hurst & Hanson y Counsel for Christa Maria, et al.

311 1/2 E. Mitchell Urban Law Institute Petoskey, Michigan 49770 Antioch School of Law 263316th Street, NW l

Washington, D. C.

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Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Ms. JoAnn Bier 204 Clinton Street Charlevoix, Michigan 49720 Thomas S. Moore Atomic Safety and Licensing Appeal Board.

. U. S. Nuclear Regulatory Commission Washington, D. C.

20555 James G. Keppler, Regional Administrator Nuclear Regulatory Commission, Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 e

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BIG ROCK POINT PLANT SEP TOPIC XV-1 EVALUATION DECREASE IN FEEDWATER TEMPERATURE I.

INTRODUCTION x

Loss of fee &ater heating can result from the closure of steam extraction line bleeder trip valves to either the.high pressure (HP) feedwater. h'sater or the intermediate pressure (IP) feedwater heater.

These valves may close as a result Feedwater of high water level on the shell side of either feedaater heater.

heating can be lost also if extraction steam is bypassed around the hea.ters.

In the second case, The first case produces a gradual cooling of the feedwat-er.

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the steam bypasses the heater and no heating. of feedwater occurs.

case, the reactor vessel receives cooler feedwater and.causes an increase in The decrease in coolant void fraction and the negative core inlet subcooling.

void reactivity coefficient result in a gradual initial increase in reactor pow er.'

The rate of power increase depends on which feedwater heater is no The~ operator will respond to any powe'r increase resulting longer functioning.

from cold feedsater b~ checking the control rod pattern and if necessary, in-maintain an acceptable power level, or shutdown the serting control rods t:

Failure of the operator to control power or water level reactor if recuire.:.

j in the feedwater heaters will cause the reactor power to increase above rated.

If cnly the Mp heater is lost, power will reach steady state above 100%, but below the high power trip setpoint of 125%.

If the IP heater or more than one heater is lost, the reactor will trip on high power thus terminating the tran-t sient.

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

t perforrance of structures, systems, and components of the facility with the ob-jective of assessing the risk to public health and safety resulting 'from operation

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. Section 50.36 of 10 CFR Part 50 requires the Technical Specifications to include

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safety limits which protect the integrity of the physical barriers which guard.

a against the uncontrolled, release of radioactivity.

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

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GDC 10 " Reactor Design" requires that the core and associated coolant, control and protection' systems be designed with appropriate mar' gin to assure that specified acceptable fuel design limits are not exceeded durUfg n'ormal operation, 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 tr.e design conditions of the reactor coolant pressure boundary are not exceeded during normal c:eration, including the effects of anticipatad operational occurrence s.

GCC 25 " Reactivity Control System Redundancy and Capability" recuires that the

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-c'. sjstems be catatle of reliac;. centrciling reactivity charges to assure that under conditions of normal operatien, including anticipated oper-ational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded.

III.

REFERENCES:

Letter from R. A. Vincent to D. M. Crutchfield dated July 15, 1981, 1.

Enclosure entitled " Plant Transient Analysis of the Big Rock Point Nuclear Reactor" m.,

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i BIG ROCK POINT PLANT

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SEP TOPIC XV-1 EVALUATION INCREASE IN FEEDWATER FLOW x

INTRODUCTION I.

Failure of the feedwater control system which causes the feedwater regulat-ing valve to open to its maximum position will permit excessive feedwater There is a gradual rise in the steam drum level and' flow to the reactor.

an increase in power because of the increase core inlei.subcooling and the negative void coefficient of reactivity.. As jt result,.the reactor will trip ' ~

A high steam drum on high power approximately 15 seconds into the., event.

.1evel alarm may not occur before'the high, power scram.

Following. reactor The oper-scram, the steam drum level will drop because of void collapse.

ator will then have approximately two minutes before a hich drum level alarm occurs, at which time the operator will place the main feedwater valve control on remote manual mode and proceed to trip the feedwater pumps and If the operator fail's to respchd to the high level alarm and the turbine.

terminate feedwater flow, water will reach the safety valves and discha ge The reactor depressurization system (RDS) and the through these valves.

cere spray system (CSS) may be needed for long-term core cooling.

l EE'!:E',. CE! TEEP l

Section 50.34 of 10CFR 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 objective of assessing the risk to public health and safety resulting from operation of the facility, including determination of th.e margins of safety during operation and transient conditions anticipated during the life

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Section 50.36 of 10 CFR Part 50 requires the Techn.ical Specifications to include safety limits which protect the intlgrity of the physical barriers which guard against the uncontroll,ed release of radioactivity.

The General Design Criteria (Appendix A to 10 CFR Part 50) establish mini-mum requirements for the principal design criteria for water-cooled reactors.

GDC 10 " Reactor Design" requires that the core and associated coolant, co'ntrol and protection systems be designed with appropriate margin to assure that spec'ified acceptable fuel design limits are not, exceeded duri'ng normal oper-ation including the effects of anticipate,d opetA}i.onal occurrences.

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 oper-ational occurrences.

l GDC 25 " Reactivity tontrol System Redundancy and Capability" requires that the reactivity control systems be capable of reliably controlling reactivity c'arces tc assure that. under conditions of nor al operation, including antici.ated operational occurrences, and with appropriate margin for mal-functions such as stuck rods, specified acceptable fuel' design limits are not exceeded.

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

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

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REVIEh GUIDELINES '

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

N 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 event as well as the ability of these systems to function as required.

V.

EVALUATION The licensee utilized the RETRAN computer code to evaluate the consequences' -

of a feedwater controller failure. The results are in increase in reactor power to a maximum value of 125.6%, a MCPR of 1.57 and maximum coolant pressure remains below the allowed (1870 psia). 'The plant initi&l operating condition was assumed' to be 102% of licensed power level and the turb.ine bypass sytem is also assumed to be functional..ln. response tokhe power incr. ease, th.e".@. f A n;..~-

.*, core? heat flux increases. to a maximum of 108% of thi.~' itial val'ue. Thel: d.'t.

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. 9*l1;' 2 in increases are not as large as those resulting from load rejection without bypass, e.g., 350.5% of initial core power and 121.4% of initial heat flux.

In Reference 2, the licensee reanalyzed the feedwater controller failure event with the turbine bypass valve failed. closed. The change in MCPR for this event was the same as for the case above since the bypass valve did not begin to open until after the minimum CPR occurred. The maximum pressure of 1416.5 psia is reached in 21.5 seconds.

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VI.

CONCLUSION As part of the SEP review of Big Rock Point, we have evaluated the licensee!s x '

analysis of a feedwater controller failure event.

Reference 1 indicates that the MCPR is 1.57 for this event and the maximum allowable re. actor coolant pressure (110% of the design pressure) would not be violated.

For the case with the turbine b,ypass valve inoperable, the MCPR and maximum reactor coolant pressure requirements are.also met. Therefore, we conclude.th.at the, results of this analysis satisfy the criteria in GD.C.10,15. and 26 as implemented by SRP Sections 15.1.1 to 15.1.4, and are acceptable..

VII.

REFERENCES -

1.

Letter from R.A. Vincent to D.M. Crutchfield, dated July 15, 1981, Enclosure entitled, " Plant Transient Analysis of the Big Rock Point Nuclear Reactor."

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Letter from R.A. Vincent to dim. Crutchfield, dated June'10,1982.

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g BIG ROCK POINT SEP TOPIC XV-1 EVALUATION INCREASE IN STEAM FLOW I.

INTRODUCTION Failure of the initial pressure regulat'or (IPR) control sys. tem can cause the turbine admission valves 'to open, allowing maximum steam flow to Primary system pressure will decrease and the resulting the turbine.

The void fraction will cause a slight decrease in reactor power.

operator must take manual action to close the turbine admission valves If control'.

until steam flow returns to the same level prior to failure.

of turbine steam flow and primary system' press'u're has not been reached

. by the time'the primary sys, tem pressure decreases to approximately 900 psig, the operator would, by procedure, trip the turbine (stop valve closed).

Failure of the operator to perform these actions will not adversely affect the reactor operation as the reactor would still be operating at a steady power level of aoproximately 100". power.

I The turbine bypass valves can also fail in such a way that the bypass i

valves open, partially or fully, causing an increase in steam flow and the same effect on the reactor as discussed previously.

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produce f

a rapid decrease in the main generator output will result.

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opposite to the effect produced by an IPR failure because steam is diverted from the turbine in this case.

Manual action would be required to close the bypass valves or close the main steam isolation ' valves which i'

The resulting effect is found to be less j

will result in a reactor scram.

i severe than IPR faiJure.

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REVIEW CRITERIA Section 50.34 of 10 CFR Par ~t 50 requires that ehch applicant for a construction' x~

permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the cbjective of assessing the risk to public health and safety resulting from the operation of the facility, including determination of the margins of safety during normal cperatiQns and transient conditions anticipated during the life of the facility.

Section 50.36 'of 10 CFR Part 50 requir'es the Te'chnical Specifications to include safety limits which protect the integrity of the physical barriers which guard against the uncontrolled release of radioactivity.

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

requirener.ts for the principal design criteria for water-cooled reacters.

~

GDC 10 "Reac cr :esign" recuires that the core and associated coolant, con-t-*

trol and protection Iystems be designed with appropriate margin to assure that specified accec:able fuel design limits are not exceeded during normal Opera-

-icr., i ci.ci g : e e##e: s of anticipated cpera-icr.ai eccur-er.ce.

GDC 15 " Reactor Coolant System Design" requires that the reactor coolant and associatec oretection systems be designed with sufficient margin to assure that the design cendi-ions of the reactor coolant pressure boundary are not exceeded during normal operation, including the effects of anticipated cperational I

occurrences.

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changes to assure that under conditions of normal operation, including anticipated operational occurrences, and w'ith appropriate margin for mal-functions such as stuck rods, specified acceptable fuel desien limits are-not exceeded.

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RELATED SAFETY TOPICS Various other SEP topics evaluate such ite$s as the reactor protection sys The effects of single failures on safe shutdown c'apability are considered

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under Topic VII-3.

TV.

REVIEW GUIDELINES 15.1.1,15.1.2,15.1'.3 and The review is conducted in accordance with SRP 15.1.2 The evaluation includes review of the analysis fcr the event and identification of the features in the plant that mitigate the consecuences of the event as The extent to well as the ability cf these systems to function as recuired.

wnicn cpera cr ac-icn is rec;uirec is also evalaatsd.

I V.

EVAll!ATION_

Failure of the turbine admission valves would result in an initial i

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l Initial sharp increase steam flow and a slight decrease in core power.

in steam flow will lead to a lower enthalpy of recirculation flow leaving the Upon reaching the core inlet the cooler water will cause the steam drum.

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previously decaying core power to increase.

The transient power peaks at approximately 99% of the original power levelb The resulting increase in core void fraction limits the small, power transient and power will level off at approximately 98% of its initial value.

Ths event is not limiting with respect to peak system pressure and minimum critical power ratio.

VI.

CONCLUSIONS As part of the SEP review for Big Rock Point, we -have evaluated the licensee's The results provided treatment of the failure of an IPR to the open position.

in Reference 1 indicate that the MCPR is 1.67 for this event and the m reactor coolant pressure (1870 psia) would not be violated. Therefore, we

~ concluded that the results are in conformance with SRP Section 15.1

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acceptabl e.

VII. " REFERENCES Vincent to D. M. Crutchfield dated July 15, 1981, 1.

Letter from R. A Enclosure entitled " Plant Transient Analysis of the Big Rock Point r n: cr."

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ML20054K053

Contents

Text

Dear Mr. VandeWalle:

SUBJECT:

Enclosure:

REFERENCES:

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ML20054K053

Text

Dear Mr. VandeWalle:

SUBJECT:

Enclosure:

REFERENCES:

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