ML20065S788

Tech Specs for Redundant Decay Heat Removal Capability, TMI-1

ML20065S788
Person / Time
Site:	Three Mile Island
Issue date:	07/31/1982
From:	Farmer F EG&G, INC.
To:	Donohew J Office of Nuclear Reactor Regulation
References
CON-FIN-A-6429 EGG-EA-5854, NUDOCS 8211020021
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hl$0 NtchaYdrlYhmcal CS/ *Mf Y EGG-EA-5854 July 1982 TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT g fM REMOVAL CAPABILITY, THREE MILE ISLAND, UNIT NO. 1 A/8f Alf7S (TMI-1) d/

F. G. Farmer U.S. Department of Energy Idaho Operations Office

• Idaho National Engineering Laboratory s

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,ygw 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-AC07-761001570 FIN No. A6429 g E G n G ,o u.

8211020021 820731 g

PDR RES 8211020021 PDR

I n E p Idaho inc FORM EG&G 396 (Rev 03-82)

INTERIM REPORT Accession No.

Report No. EGG-EA-5854 Contract Program or Project

Title:

Selected Operating Reactors Issues Program (III)

Subject of this Document:

Technical Specifications for Redundant Decay Heat Removal Capability, Three Mile Island, Unit No. 1 (TMI-1)

Type of Document: ,

Informal Report Author (s):

F. G. Farmer Date of Document:

July 1982 Responsible NRC Individual and NRC Office or Division:

J. N. Donohed, Division of Licensing 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 i

Prepared for the

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

Under DOE Contract No. DE-AC07 761D01570 i

• NRC FIN No. A6429 l

lNTERIM REPORT l

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TECHNICAL SPECIFICATI0NS FOR REDUNDANT DECAY HEAT REMOVAL CAPABILITY THREE MILE ISLAND, UNIT NO. 1 (TMI-1)~

July 1982

.F. G. Farmer Reliability and Statistics Branch Engineering Analysis Division EG&G Idaho, Inc.

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. Docket No. 50-289 TAC No. 42127 d

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m ABSTRACT This report reviews the Three Mile Island, Unit No. 1 (TMI-1),

technical specification requirements for redundancy in decay heat removal capability in all modes of operation. Applicable portions of the TMI-l technical specifications were compared with the model technical specifica-tions for Babcock & Wilcox pressurized water reactors. The TMI-l technical specification requirements do not agree with those of the model technical specifications, and some revision to them may be necessary.

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 20-19-10-11, FIN No. A6429.

Selected Operating Reactors Issues Program (III)

FIN No. A6429 ii u- ,.- -

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' CONTENTS 1.0 . INTRODUCTION .................................................... 1

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

. 3.l' Startup and Power Operations .............................. 2 3.2 Hot Standby ............................................... 2 3.3 Shutdown .................................................. 2 3.4- Refueling ................................................. 2

4.0 CONCLUSION

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5.0 REFERENCES

...................................................... 3 APPENDIX A--NRC MODEL TECHNICAL SPECIFICATION ........................ 5' O

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s TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT REMOVAL CAPABILITY ,

THREE MILE-ISLAND, UNIT NO. 1 (TMI-1) 1.0 JNTRODUCTION.

,' A number of events have occurred at operating PWR facilties where decay heat removal capability has been seriously degraded due to inadequate administrative controls during shutdown modes of One of these 3 events,describedinIEInformationNotice80-20,9peration. occurred at the 0#vis' Besse, Unit No.1 plant, on April 19,.1980. In IE Bulletin 80-12,2 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 bul-letin was -to effect imediate 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 propose technical specifications (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 solu-tion of the concern and include an appropriate safety analysis as a basis; and submit the proposed TS with the basis by October 11,1980.

GPU Nuclear has proposed that'the existing TMI-l TS provide adequat redundancyindecayheatremovalcapabilityanddonotrequirerevision.g 2.0 REVIEW CRITERIA The review criteria for this task are contained in the June 11, 1980 letter f rom the NRC to all PWR licensees. The NRC provided the model tech-nical specifications (MTS)5 which identify the normal required redundant coolant systems and the required actions when redundant systems are not available for a typical two-loop plant ( Appendix A). The general review criteria are:

1. Two independent methods for decay heat removal are required in the plant TS for each operating mode.

2. Periodic surveillance requirements should insure the operability of the systems.

3.0 DISCUSSION AND EVALUATION Three Mile Island, Unit 1, is a two-loop, Babcock and Wilcox (B&W) PWR plant. .The TMI-l TS dif fer in format from the Nuclear Regulatory Com-missions (NRC) MTS for B&W PWRs. Limiting conditions for operation start with Section 3 of the TMI-1 TS. Section 3 does not delineate the limiting conditions by applicable modes as in the MTS. The modes identified in this

.- report are based on definitions found in Section 1 of TMI-l TS. Similarly, Section 4 of the TMI-l TS does not. define the surveillance requirements based on applicable modes. -

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'3.l' Startup and Power Operation--Modes 1 and 2 (Toperating}

The TMI-1 TS require that both reactor coolant loops and both reactor coolant pumps;i_n each ioop be operating at startup or at full power.6 With one coolant pump not in operation, the TMI-l TS require a reduction to

-75% of full operating power, in agreement with the MTS. Loss of a 'colant pump automatically reduces setpoints for nuclear overpower based on flow and imbalance and nuclear overpower based on pump monitors; the TMI-l TS do -

not require reduction of the nuclear overpower setpoint (power independent of.any other parameter). ,

The TMI-l TS do not require verification of pump and loop operation on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis, and do not require verification of the reactor protection system setpoint changes within four hours.

3.2 Hot Standby--Mode 3 (Tave > 525 F and Keff = 1.00)

The TMI-l TS permit operation with one reactor coolant pump in each loop idle for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; if the reactor is not returned to an acceptable operating pump / loop condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor must be in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The TS also prohibit boron concentration reduc-tion unless at least one reactor coolant or decay heat removal pump is circulating reactor coolant.

The TMI-1 TS do not include the surveillance requirements of MTS 4.4.1.2.1 and 4.4.1.2.2.

3.3 Shutdown--Mode 4 (TAVG < 525 F), Mode 5 (TAVG 1200*F), K eff 1 0.99 The TMI-l TS require that both steam generators be operable, as well as the turbine-driven emergency feedwater pump and two half-sized motor-driven emergency feedwater pumps, when TAVG > 250 F; there is, however, no requirement that at least one of the loops listed in MTS 3.4.1.3 be in operation, nor are operability requirements specified for TAVG 1 250 F.

The TMI-l TS do not contain requirements for surveillance of the oper-ability of the DHR, reactor coolant pumps or steam generator loops in i Modes 4 and 5.

3.4 Refueling Operations--Mode 6 The TMI-l TS require at least one DHR loop operating in Mode 6; there is no requirement that two independent DHR loops be operable in Mode 6 with less than 23 f eet of water above the top of irradiated fuel assemblies.

Neither is there a requirement to close all containment penetrations pro-viding direct access from containment atmosphere to the outside atmosphere within four hours of loss of DHR loop operation. -

The TMI-l TS do not require surveillance of DHR loops as specifiea in ,

MTS 4.9.8.1 and 4.9.8.2.

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4.0 CONCLUSION

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'The TMI-1.TS do not agree with the MTS in the following areas:

-(l) They do not follow the MTS format in' defining the limiting condi-tions by applicable modes, 3 (2) They differ from the MTS 'in Modes 1 and 2 by not requir:ing reduc-tion of nuclear power monitor setpoints when operating with less than four coolant pumps, (3) They do not require having at least one. coolant loop in operation in Modes 4 and 5, (4) They do not_ require two DHR loops operable in Mode 6 with less than 23 feet of water above irradiated fuel assemblies, (5) They do not require closing of containment penetrations in Mode 6 witi less than one DHR loop operating, (6) They do not reouire the-surveillance specified in the MTS in Modes'1 through 6.

5.0 REFERENCES

1. NRC 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. GPU Nuclear letter, H. D. Hukill to NRC, J. F. Stolz, " Decay Heat Removal (DHR) Technical Specifications," dated January 26, 1982.

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

6. Technical Specifications for Three Mile Island, Unit 1, Amendment 41 to the Final Safety Analysis Report, dated April 16, 1973.

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

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3/4.4 -REACTOR COOLANT SYSTEM.

3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 Both reactor coolant loops.and both reactor coolant pumps in each loop.shall'be in operation.

APPLICABILITY: MODES 1 and 2.*

ACTION:

With one. reactor coolant pump not in operation, STARTUP and POWER OPERATION may be initiated and may proceed provided THERMAL POWER is restricted to less than ( )% of RATED THERMAL POWER and within 4~ hours the setpoints for the following trips have been reduced to the values specified in Specification 2.2.1 for operation with three reactor coolant pumps operating:

1. (NuclearOverpower).

2. (Nuclear Overpower based on RCS flow and AXIAL POWER IMBALANCE).

3. (Nuclear-Overpower based on pump monitors).

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 />.

4.4.1.2 The Reactor Protective Instrumentation channels specified in the applicable ACTION statement above shall be. verified to have had their trip setpoints changed to the values specified in Specification 2.2.1 for the applicable number of reactor coolant pumps operating either:

a. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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.

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 . The reactor coolant loops listed below shall be OPERABLE:

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1. Reactor Coolant Loop (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

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 />.

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b. With no reactor coolant loop in operation, suspend all operations r

involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate action to return the i

required coolant loop to operation.

i SURVEILLANCE REQUIREMENT i

l 4.4.1.2.1 At least the above recuired reactor coolant pumps, if not in

' oper6 tion, shall be determined te 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 />.

l 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 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 at least one associated reactor coolant pump,

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

3. Decay Heat Removal Loop (A),*

4. Decay Heat Removal Loop (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 <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.

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• 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 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 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 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 ( )%.

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 L. .

LIMITING CONDITION FOR OPERATION 3.9.8.1 ~ At lea',t 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 baron 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b. The DHR loop may be removed from operation for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 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 <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 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.

I' ACTION:

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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.

• '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 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 reacter 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 reactor 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, singleffailure con-siderations require that two loops be OPERABLE.

In MODES 4 and 5, a single reactor coolaat 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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-3/4.9.81 DECAY HEAT REMOVAL' AND COOLANT CIRCULATION The requirement that at least one DHR loop be in operation ensures w .

~that-(1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor 0

~ during the REFUELING MODE, and sufficient(2) pressure vessel-coolant. below 140 circulation is F as required main-tained through the.reactoricore to minimize the effect of a boron dilution ,

incident and prevent boron stratification. ,.

The-requirement to-have two DHR loops'0PERABLE when there is less than

23 feet of water-above the core ensures that a single failure of the 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.

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