ML20041D725

Tech Specs for Redundant Decay Heat Removal Capability, Donald C Cook Nuclear Plant,Units 1 & 2, Preliminary Technical Evaluation Rept

ML20041D725
Person / Time
Site:	Cook
Issue date:	01/31/1982
From:	Steverson J EG&G, INC.
To:	Donohew J Office of Nuclear Reactor Regulation
References
CON-FIN-A-6429 EGG-EA-5746, EGG-EA-5746-DFT, NUDOCS 8203090029
Download: ML20041D725 (16)

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i EGG-EA-5746 JANUARY 1982 fdA TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT ////

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.- A, This is an informal report intended for use as a preliminary or working document Prepared for the U.S. Nuclear Regulatory Commission Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6429 E G n G ,s.n.

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FOPW EG4G M8 c- . n INTERIM REPORT Accession No.

Report No. EGG-EA-5746 Contract Program or Project

Title:

Selected Operating Reactor Issues Program (III)

Subject of this Document Technical Specifications for Redundant Decay Heat Removal Capability, Donald C. Cook Nuclear Plant, Unit Nos. I and 2 Type of Document:

Technical Evaluation Report Author (s):

J. A. Steverson Date of Document:

January 1982 Responsible NRC/ DOE Individual and NRC/ DOE Office or Division:

J. N. Donohew, Division of Licensing, NRC This document was prepared primarily for preliminary or internat use. it has not received full review and approval. Since there may be substantive changes, this document should not be considered final.

EG&G Idaho, Inc.

Idaho Falls ldaho 83415 l

l Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.

Under DOE Contract No. DE AC07 761D01570 l NRC FIN No. A6429 INTERIM REPORT f

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. TECHNICAL EVALUATION REPORT TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT REMOVAL CAPABILITY DONALD C. COOK NUCLEAR PLANT, UNIT N05. 1 AND 2 Docket Nos. 50-315 and 50-316 January 1982 J. A. Steverson Reliability and Statistics Branch Engineering Analysis Division

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EG&G Idaho, Inc.

Draft 1/13/82 TAC No. 42093 and 42094

ABSTRACT This report reviews the Donald C. Cook Nuclear Plant, Unit Nos.1

- and 2, proposed technical specifications for redundancy in decay heat removal capability in all modes of operation.

FOREWORD This report is supplied as part of the " Selected Operating Reactor Issues Program (III)" being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Licensing, by EG&G Idaho, Inc., Reliability and Statistics Branch.

The U.S. Nuclear Regulatory Commission funded the work under the authorization, B&R 20 19 02 06, FIN No. A6429.

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TECHNICAL EVALUATION REPORT TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT REMOVAL CAPABILITY DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2

1.0 INTRODUCTION

A number of events have occurred at operating PWR facilities where decay heat removal capability has been seriously degraded due to inadequate administrative controls during shutdown modes of operation. One of these events, described in IE Information Notice 80-20,I occurred at the Davis-2 Besse, Unit No.1 plant on April 19, 1980. In IE Bullet 1n 80-12 dated May 9,1980, licensees were requested to immediately implement administra-tive controls which would ensure that proper means are available to provide 1

redundart methods of decay heat removal. While the function of the bul-letin was to effect immediate action with regard to this problem, the NRC considered it necessary that an amendment of each license be made to pro-vide for permanent long-term assurance that redundancy in decay heat removal capability will be maintained. By letter dated June 11, 1980,3 all PWR licensees were requested to: 1) propose technical specification (TS) changes that provide for redundancy in decay heat removal capability in all l

modes of operation, 2) use the NRC model TS which provide an acceptable solution of the concern and include appropriate safety analyses as bases, and 3) submit the proposed TS with the bases by October 11, 1980.

Indiana & Michigan Electric Company, New York, New York, submitted pro-posed revisions for decay heat removal to their technical specifications i -

for Donald C. Cook Nuclear Plant, Unit Nos.1 and 2,4 on August 14, 1981.

The following discussion evaluates the proposed TS and notes any differences existing between them and the model TS provided by the NRC ( Appendix A).

The requirements are compared for equivalent modes of operation.

2.0 DISCUSSION D. C. Cook, Units 1 and 2 are four-loop Westinghouse PWR plants.

Specific sections of the Westinghouse Standard Technical Specifications 5 that apply to this task are:

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3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION 3/4.9 REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The proposed D. C. Cook technical specifications are in very close agreement with the model TS provided by the NRC. Surveillance Require-ment 4.9.8.1, however, states that at least one residual heat removal loop must be verified to be in operation and circulating reactor coolant at a ,

flow rate of greater than or equal to 2800 gpm at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The proposed TS require the flow rate of at least 3000 gpm to be verified at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. All other Limiting Conditions and Surveil-lance Requirements in the D. C. Cook proposed TS are in complete agreement with those in the NRC model TS.

3.0 REFERENCES

1. NRC IE Information Notice 80-20, May 8,1980.

2. NRC IE Bulletin 80-12, May 9, 1980.

3. NRC letter, D. G. Eisenhut, To All Operating Pressurized Water Reac-tors (PWR's), dated June 11, 1980.

4. Indiana & Michigan Electric Co. letter, G. P. Maloney to H. R. Denton, dated August 14, 1981.

5. Standard Technical Specifications for Westinghouse Pressurized Water Reactors, NUREG-0452-Rev. 3, Fall 1980.

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APPENDIX A MODEL TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT REMOVAL FOR WESTINGHOUSE PRESSURIZED WATER REACTORS (PWR's) 3 i

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 All reactor coolant loops shall be in operation.

APPLICABILITY: MODES 1 and 2.*

ACTION:

With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

SURVEILLANCE REQUIREMENT 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• See Special Test Exception 3.10.4.

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REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 a. At least two of the reactor coolant loops listed below shall be OPERABLE:

1. Reactor Coolant Loop (A) and its associated steam -

generator ar.d reactor coolant pump,

2. Reactor Coolant loop (B) and its associated steam

• generator and reactor coolant pump,

3. Reactor Coolant Loop (C) and its associated steam generator and reactor coolant pump, 4 Reactor Coolant Loop (D) and its associated steam generator and reactor coolant pump,

b. At least one of the above coolant loops shall be in operation.*

APPLICABILITY: MODE 3 ACTION:

a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 100F below saturation temperature.

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REACTOR COOLANT SYSTEM

b. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENT

. 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4:1.2.2 At least one cooling loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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r REACTOR COOLANT SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the coolant loops listed below shall be OPERABLE:

1. Reactor Coolant Loop (A) and its associated steam gen- -

erator and reactor coolant pump,*

2. Reactor Coolant Loop (B) and its associated steam gen-erator and reactor coolant pump,*

3. Reactor Coolant Loop (C) and its associated steam gen-erator and reactor coolant pump,*

4. Reactor Coolant Loop (D) and its associated steam gen-erator and reactor coolant pump,*

5. Residual Heat Removal Loop (A),**

6. Residual Heat Removal Loop (B).**

b. At least one of the above coolant loops shall be in operation.***

• A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to (275)0F unless 1) the pressurizer water volume is less than cubic feet or 2) the secondary water temperature of each steam generator is less than OF above each of the RCS cold leg temperatures.

• • The normal or emergency power source may be inoperable in MODE 5.

• • • All reactor coolant pumps and decay heat removal pumps may be de-energized for up to I hour provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 0

2) core outlet temperature is maintained at least 10 F below saturation temperature.

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REACTOR COOLANT SYSTEM APPLICABILITY: MODES 4 and 5.

ACTION:

a. With less than the above required loops OPERABLE, innediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

b. With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENT 4.4.1.3.1 The required residual heat removal loop (s) shall be determined OPERABLE per Specification 4.0.5.

4.4.1.3.2 The required reactor coolant pump (s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.3.3 The required steam generator (s) shall be determined OPERABLE by verifying secondary side level to be greater than or equal to ( )% at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.1.3.4 At least one coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ALL WATER LEVELS LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be in .

opera tion.

APPLICABILITY: MODE 6 ACTION:

a. With less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all contain.nent penetrations providing direct access from the con-tainment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b. The residual heat removal loop may be removed from operation for up to i hour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel (hot) legs.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENT 4.9.8.1 At least one residual heat removal loop shall be verified to be in operation and circulating reactor coolant at a flow rate of greater than or .

equal to (2800) gpm at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent Residual Heat Removal (RHR) loops shall be OPERABLE.*

APPLICABILITY: MODE 6 when the water level above the top of the irradiated fuel assemblies seated within the reactor pressure vessel is less than 23 feet.

ACTION:

a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible,

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENT 4.9.8.2 The required Residual Heat Removal loops shall be determined 0FCPABLE per Specification 4.0.5.

I . _ _

• The normal or emergency power source may be inoperable for each RHR loop.

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3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not .

in operation this specification requires that the plant be in at least HOT STANDBY within I hour. ,

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure con-siderations require that two loops be OPERABLE.

In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to (275)0F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the .

limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to ' '

expand into, or (2) by restricting starting of the RCPs to when the secon-dary water temperature of each steam generator is less than ( )0F above each of the RCS cold leg temperatures.

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REFUELING OPERATIONS BASES 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that (1) sufficient cooling capacity is available to

. remove decay heat and maintain the water in the reactor pressure vessel below 140 F as required during the REFUELING MODE, and (2) sufficient cool-ant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the core ensures that a single failure of the oper-ating RHR loop will not result in a complete loss of residual heat removal capab ility. With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling. Thus, in the event of a f ailure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

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