Text

Tech Specs for Redundant Decay Heat Removal Capability, AR Nuclear One,Unit 1, Technical Evaluation Rept
	ML20050A294
Person / Time
Site:	Arkansas Nuclear
Issue date:	02/28/1982
From:	Steverson J; EG&G, INC.
To:	Donohew J; Office of Nuclear Reactor Regulation
References
	CON-FIN-A-6429 EGG-EA-5711, NUDOCS 8204010027
	Download: ML20050A294 (17)
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_ www m em l 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-AC07-761D01570 FIN No. A6429 g p EGnG ,s.no 0204010027 820229 PDR RES PDR 8204010027

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FORM EG&G 398 in.,ii.n INTERIM REPORT Accession No.

Report No. EGG-EA-5711

Contract Program or Project

Title:

Selected Operating Reactors Issues Program (III) e Subject of this Document:

Technical Specifications for Redundant Decay Heat Removal Capability, Arkansas Nuclear One, Unit No. 1 Type of Document:

Technical Evaluation Report Author (s):

J. A. Steverson Date of Document:

February 1982 Responsible NRCIDOE Individual and NRCIDOE Office or Division:

J. N. Donohew, Division of Licensing, NRC This document was prepared primarily for preliminary or internal 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, Idaho 83415 Prepared for the

U.S. Nuclear Regulatory Commission

! Washington, D.C.

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

0013j o

TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT REMOVAL CAPABILITY ARKANSAS NUCLEAR ONE, UNIT N0. 1 Docket No. 50-313 February 1982 J. A. Steverson Reliability and Statistics Branch Engineering Analysis Division EG&G Idaho, Inc.

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TAC No. 42120

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i ABSTRACT l This report reviews the Arkansas Nuclear One, Unit No.1 proposed technical specification requirements 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 Connission, 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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3 CONTENTS

1.0 INTRODUCTION

.................................................... l-I 2.0 REVIEW CRITERIA ................................................. I 3.0 DISCUSSION AND EVALUATION ....................................... 1

, , 3.1 Startup and Power Operation--Modes 1 and 2 ................ 2 3.2 Hot Standby--Mode 3 ....................................... 2

3.3 Hot and Cold Shutdown--Modes 4 and 5 . . . . . . . . . . . . . . . . . . . . . . 3 l 3.4 Refueling--Mode 6 ......................................... 3 9

4.0 CONCLUSION

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5.0 REFERENCES

...................................................... 4 APPENDIX A--MODEL TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT REMOVAL FOR BABC0CK AND WILC0X PRESSURIZED WATER REACTORS .................................................. 5 l

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TECHNICAL EVALUATION REPORT TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT REMOVAL CAPABILITY ARKANSAS NUCLEAR ONE, UNIT NO. I

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 9peration. One of these

, events,describedinIEInformationNotice80-20,'occurredattgeDavis-Besse, Unit No.1 plant on April 19, 1980. In IE Bulletin 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 redundant methods of decay heat removal. While the function of the bulle-tin was to effect irrmediate action with regard to this problem, the NRC con-sidered it necessary that an amendment of each license be made to provide forpermanentlongtermassurancethatredundancyindecaygeatremovalcap-ability will be maintained. By letter dated June 11, 1980, all PWR licensees were requested to propose technical specification (TS) changes that provide for redundancy in decay heat removal capability in all modes of operation; use the NRC model TS which provide an acceptable solution of the concern and include an appropriate safety analysis as a basis; and submit tne proposed TS with the basis by October 11, 1980.

Arkansas Power & Light (AP&L), Little Rock, Arkansas, submitteo pro-posed revisions for decay heat removal to their technical specifications (TS) for Arkansas Nuclear One, Unit 1,4 on October 31, 1980.

2.0 REVIEW CRITERIA The review criteria for this task are contained in the June 11, 1980 letter from the NRC to all PWR licensees. The NRC provided the model tech-nical specifications (MTS) which identify the normal required redundant coolant system and the required action when redundant systems are not avail-able for a typical two loop plant (Appendix A). The purpose of this report is to review the licensee's proposed TS and note any differences between them and the model TS as provided by the NRC.

3.0 DISCUSSION AND EVALUATION Arkansas Nuclear One, Unit 1 (AN0-1) is a two coolant loop Babcock &

Wilcox (B&W) PWR plant. The following discussion presents an evaluation of o the proposed technical specifications submitted by AP&L for redundant decay heat removal as requested by the NRC. Because ANO-l's proposed TS are not in the NRC MTS format, wording and organization are quite different. This i evaluation compares the ANO-1 proposed TS to the NRC MTS requirements dur-ing equivalent operating conditions; the conditions defining each of the followgngmodesarethosegivenbytheB&WStandardTechnicalSpecifica-tions.

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3.1 Startup and Power Operation--Modes 1 and 2 (K >0.99, Rated Thermal Power >5%, T >305 F)

The NRC model TS require both coolant loops and both reactor coolant pumps in each loop to be in operation. With one of the four coolant pumps not in operation, STARTUP and POWER OPERATION may be initiated and may proceed provided thermal power is restricted. Within four hours of losing one pump, the setpoints for the following trips are required to be reduced:

1) Nuclear Overpower, 2) Nuclear Overpower based on RCS flow and AXIAL

POWER IMBALANCE, and 3) Nuclear Overpower based on pump monitors.

The proposed ANO-1 TS refer to Table 15.2.3-1 of the techn e ficationsforrequiredpumpcombinationsatgivenpowerlevels.gcalspeci- Four pumps are required at full power. With the loss of one pump, setpoints for

1) Nuclear Overpower and 2) Nuclear Overpower based on RCS flow and AXIAL POWER IMBALANCE are reduced as required; however, no time limit (as 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) is specified. The table indicates that resetting the trip values for Nuclear Overpower based on pump monitors is not applicable with three pumps in operation.

i i Periodic surveillance is required by the NRC MTS to ensure that the limiting conditions discussed above are satisfied. However, the MTS also j require verification that the trip setpoints be reduced either (a) within

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after switching to a different pump combination if the switch is

made while operating, or (b) prior to reactor criticality if the switch is made while shutdown. This requirement is not included in the proposed TS.

J 3.2 Hot Standby--Mode 3 (Keff <0.99, Rated Thermal Power = 0%, T avg #305 - F)

The NRC model TS require Reactor Coolant Loop (A), Reactor Coolant Loop (B), and at least one associated reactor coolant pump in each loop to be operable in mode 3. At least one of the coolant loops and an associated pump must be in operation.a With less than the above reactor coolant loops operable, the MTS require the loop (s) be restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise the reactor must be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . With no reactor coolant loop in operation, all operations i involving a reduction in boron concentration must be suspended and immediate action must be taken to return the required loop to operation.

The licensee's proposed TS, (having T av >280 F for mode 3), satisfy

j. the conditions required by the NRC MTS. No ktatement is made, however, con-

cerning the action taken when there are no coolant loops in operation, i.e.

suspending all operations involving .

4 Also, the special condition in note (a) reduction a below, in discussed is not boron concentration.

in the pro-posed TS.

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a. All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 10 F below saturation temperature.

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The NRC MTS surveillance requirements assuring operation and oper-ability are met in ANO-l's proposed TS.

3.3 Hot and Cold Shutdown--Modes 4'and 5 (Keff <0.99, Rated Thermal Power = 0%, 200 F <T,yg < 305*F and Tavg 1200 F)

The NRC MTS require in modes 4 and 5, at least two of the following coolant loops to be operable: Reactor Coolant Loop (A), Reactor Coolant 4

9 Loop (B), (including their associated steam generators and at least one associated reactor coolant pump), Decay Heat Removal Loop (A)a and Loop (B).a At least one of the above coolant loops must be in 7 operation.b With less than the required coolant loops operable, the MTS require immediate action to return the loop (s) to OPERABLE status as soon as possible or the plant must be in COLD SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months . With no coolant loops operating, all operations involving a reduction in boron l concentration of the Reactor Coolant System must be suspended.

The licensee's proposed TS meet all the NRC MTS operating and surveillance requirements for modes 4 and 5.

3.4 Refueling--Mode 6 (Keff <0.95, Rated Thermal Power = 0%, T 1 140 F) l During refueling operations, the MTS require at least one decay heat removal (DHR) loop to be in operation. With less than one DHR loop in i

operation, all operations involving an increase in the reactor decay heat load or a reduction in boron concentration must be suspended. The MTS also require all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere be closed within four hours. The DHR loop may, however, be removed from operation for up to I hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel (hot) legs.

4 During refueling when the water level above the top of the irradiated i fuel assemblies seated within the reactor pressure vessel is less than 23 feet, the MTS require two independent DHR loops to be operable.c With less than two loops operable, immediate corrective action must be taken to return the required loops to OPERABLE status as soon as possible, o a. The normal or emergency power source may be inoperable in MODE 5.

b. All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 10 F below saturation temperature.

c. The normal or emergency power source may be inoperable for each DHR loop.

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The proposed TS meet the above requirements. The NRC MTS, however.

require surveillance once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to verify that at least one DHR loop is operating and circulating reactor coolant at a flow rate of equal to or greater than 2800 gpm. ANO-l's proprosed TS do not discuss the flow rate of the DHR loop.

4.0 CONCLUSION

An evaluation of the proposed TS for Arkansas Nuclear One, Unit No.1, g indicates they are in general agreement with the NRC model technical speci-fications for redundant decay heat removal. The following differences were noted and discussed in previous sections of this report: ,

l. With the loss of one pump in startup and power operation, ANO-1 TS do not specify a time limit for which trip setpoints must be reduced.

2. With the loss of one pump in startup and power operation, ANO-1 TS do not require a reduction of the trip setpoint for Nuclear Over-power based on pump monitors.

3. When there are no coolant loops in operation in hot standby, ANO-1 TS do not require that all operations involving a reduction in boron concentration must be suspended.

4. During refueling, ANO-l TS do not require the flow rate of the reactor coolant in the DHR loop to be equal to or greater than 2800 gpm.

5.0 REFERENCES

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

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

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

4. AP&L letter, W. Cavanaugh to NRC, D. G. Eisenhut, dated Octobei 31, 1980.

l l S. Standard Technical Specifications for Babcock and Wilcox Pressurized l Water Reactors, NUREG-0103-Rev. 3, July 1979.

l l 6. Arkansas Nuclear One, Unit 1, FSAR, Volume II, ( Arkansas Power & Light .

I Co. , L ittle Rock , Arkansas) 1971.

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APPENDIX A MODEL TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT REMOVAL FOR BABC0CK & WILC0X PRESSURIZED WATER REACTORS (PWR's) 6 5

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APPLICABILITY: MODES 1 and 2.*

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With orie" reactor coolut pump rict in operation, STARTUP and POWER OPERATION inay b's initiated and'may proUed provided THERMAL POWER is restricted to

-less than ( )% of RATED THERMAL POWER a'nd within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the setpoints for tne following trips have been reduced! to the'v61ues specified :in j Specification 2.2.-1 for operation with thr@. reactor coolant pumps oppfating:

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5 ,, SURVEILLANCE REQUIREMENT -

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. 4.4.1.1 The abdve required reactor coolant loops sh'all be y'erified to be in operation and circulating react 6r coolant-at,least once per~12 hours.

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4.4.1.2 The Reactor Protective Indrumentation chenpels specified in the

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applicable ACriON statement above sh'all be verifi?.d Eo have had their trip l

_.. setcoints changed (o the~ values specifiedzin Specification 2.2.1 for the applicable number ~of reactor coolant pumps operating' dither: ,

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_% a. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months af ter switching ,to,a dif ferent pump lctmbination if

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REACTOR C00LArti SYSTEM tiOT STAfiDBY ,

LIMITif4G C0f4DIT10N FOR OPERATI0ri 3.4.1.2 a. The reactor coolant loops listed t,elow shall be OPERABLE: '

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1. Reactor Coolant Lcen (A) and its associated reactor coolant pump, -

2. Reactor Coolant Loop (B) and its associated reactor '

coolant pump,

b. At least one of the above Reactor Coolant Loops shall be in operation.*

APPLICABILITY: MODE 3 ACTIOft:

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

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

SURVEILLAf4CE REQUIREMENT 4.4.1.2.1 At least the above required reactor coclant 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

All reactor coolant 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 (2) core outlet temperature is maintained at least 100F below saturation temperature.

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REACTOR COOLANT SYSTEM SHUT 00WN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the coolant loops listed below shall be I g OPERABLE:

~l . Reactor Coolant Loop (A) and its associated steam gen-3 erator and at least one associated reactor coolant pump,

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

3. Decay Heat Removal Loop (A),*

4. DecayHeatRemovalLoop(B),*

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

APPLICABILITY: MODES 4 and 5.

ACTION:

a. With less than the above required loops OPERABLE, immediately 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 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ,

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.

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

l ** 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 s would cause dilution of the reactor coolant system boron c.oncentration, and

! (2) core outlet temperature is maintained at least 10 0F below saturation temperature.

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REACTOR COOLANT SYSTEM 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 opert. ion, shall a 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 ( )%.

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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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 (DHR) loop shall be in

, operation.

APPLICABILITY: MODE 6 ACTION:

a. With less than one DHR 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 containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. The DHR loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 DHR 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent DHR 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 DHR loops OPERABLE, immediately initiate corrective action to return the required 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 DHR loops shall be determined OPERABLE per Specifica-tion 4.0.5.

t The normal or emergency power source may be inoperable for each DHR loop.

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3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION o The plant is designed to operate with both reactor coolant loops in operation, and maintain DNBR above (1.32/1.30) during all normal operations and anticipated transients. With one reactor coolant pump not in operation

in one loop, THERMAL POWER is restricted by the Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE and the Nuclear Overpower Based on Pump Monitors trip, ensuring that the DNBR will be maintained above (1.32/1.30) at the maximum possible THERMAL POWER for the number of r aactor coolant pumps in operation or the local quality at the point of minimum DNBR equal to (22/15)%, whichever is more restrictive.

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 DHR 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 DHR loops to be OPERABLE.

The operation of one Reactor Coolant Pump or one DHR 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.

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REFUELING OPERATIONS BASES 3/4.9.8 ffCAY HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one DHR loop be in operation ensures ,

that (l) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 1400F as required during the REFUELING MODE, and (2) sufficient coolant circulation is main- g tained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.

The requirement to have two DHR loops OPERABLE <.ii6r thera is less than 23 feet of water above the core ensures that a single failure of +he oper-ating DHR loop will not result in a complete loss of decay heat removal capability. 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 failure of the operating DHR loop, adequate time is provided to initiate emergency procedures to cool the core, t

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ML20050A294

Contents

Text

Title:

1.0 INTRODUCTION

4.0 CONCLUSION

5.0 REFERENCES

1.0 INTRODUCTION

4.0 CONCLUSION

5.0 REFERENCES

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ML20050A294

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Title:

1.0 INTRODUCTION

4.0 CONCLUSION

5.0 REFERENCES

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4.0 CONCLUSION

5.0 REFERENCES

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