ML20054M879

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Tech Specs for Redundant Decay Heat Removal Capability, Tmi,Unit 1 (TMI-1), Preliminary Rept
ML20054M879
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 04/30/1982
From: Farmer F
EG&G, INC.
To: Donohew J
Office of Nuclear Reactor Regulation
References
CON-FIN-A-6429 EGG-EA-5854, NUDOCS 8207150080
Download: ML20054M879 (19)
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EGG-EA-5854 April 1982 TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT 480$

. REMOVAL CAPABILITY, THREE MILE ISLAND, UNIT NO. 1 X/6/C 0f*/ S (TMI-1) g h$$ feGPato an $/ + cal f:rnflanet 8

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. s . .~ + achhh 7 e 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 h EGsG,s u.

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INTERIM REPORT Accession No.

  • Report No. EGG-EA-5854 o Centract 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 D:te of Document:

April 1982 R:sponsible NRC Individual and NRC Office or Division:

J. N. Donohew, Division of Licensing This document was prepared primari!y 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 e

Prepared for the U.S. Nuclear Regulatory Commission

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

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TECHNICAL SPECIFICATIONS FOR RED'JNDANT DECAY HEAT REMOVAL CAPABILITY THREE MILE ISLAND, UNIT N0. 1 (TMI-1)

Docket No. 50-289 April 1982 F. G. Farmer Reliability and Statistics Branch Engineering Analysis Division EG&G Idaho, Inc.

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TAC No. 42127

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ABSTRACT In response to the D. G. Eisenhut letter dated June 11, 1980, GPU Nuclear replied that existing TMI-l Technical Specifications are adequate and no new Technical Specifications are necessary to increas'e aware' ness of Decay lieat Removal system availability.

FOREWORD ,,

This report is supplied as part of the " Selected Operating Reactors' 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. '

,j The U.S. Nuclear Regulatory Commissioni f,unded the work under the authorization, B&R 20-19-01-06, FIN No. A6429.

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r CONTENTS

1.0 INTRODUCTION

.................................................... I 2.0 REVIEW CRITERIA ................................................. I O 3.0 DI SCUSSI ON AND E VALUAT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.1 Startup and Power Operations .............................. 2 3.2 Hot Standby ............................................... 2 3.3 Shutdown .................................................. 3 3.4 Refueling ................................................. 3

4.0 CONCLUSION

S ..................................................... 4

5.0 REFERENCES

...................................................... 4 APPEtiDIX A--hRC MODEL TECHNICAL 3PECIFICATION ........................ 5 s ,

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TECHNICAL SPECIFICATIONS _ FtM REDUNDANT DECAY HEAT REMOVR CAPABILITO <

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'P!REE MILE ISLAND, U3!1' NO.1 (TMI-1) .V q ',' 7 T' '

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  • .t 1.0 INTRODUCf!ON .s

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3 A number of events a/e'occurrejatioperatingPWRfac11tiespere? ,;

decay heat removal capab'ility,has hcen seriously degrd&.d os duey.o inadequate a

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s administrativeconCrulsduringshutdown,modesofoperation. One of these y .. ,4 'f -,

events, described in IE Information Notice 80-20, occurred at the Dapis- , / f ' ' c '..

Besse, Unit No. I plari, on April 19, 1980. In IE Bulletin -12,2 dated fj,# 5 May9,1980,licenseeswererequestedtoimmediatelyimplemenjadministra-tivecontrolswhichwouldensurethatpropermeansareavailhaletoprovide redundant methods of decay heat removal. While the function

  • of they bul-letin wan to effect immediate action with regard to thii problem, f,he NRC J

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considered it necescary that an amendment of each license be made to pro ,

vide for permanent long-term assurance tnat redundancy in decay heat removal capability will be maintained. ByletterdatddJune 11,-1980,3 all PWR licensees were requested to propose Technical Specifications (TS) > r changes that provide for redundancy in decay hect removal capability in all modes of operation; use the NRC model TS which provide an acce3 table solu-

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tion of the concern and include an appropriate safety analysis as a basis; ,

and submit the proposed TS with the basis by October' 1, 1950. >

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GPU Nuclear has proposed that the existing TM -P TS provide adequate f redundancyindecayheatremovalcapabilityanddohotrequirerevision.4 +

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, 2.0 REVIEW CRITERIA .

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ThereviewcriteriaforthistaskarecontainedintheJcNe 11, 1980 letter from the NRC to all PWR licensees. The NRC'provideo the model tech-

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nical specifications (MTS)5 which identify the normal required redundant coolant systems and the required actions when redundant systems are not e

available for a typical two-loop plant (Appendix A). The general review

( criteria are:

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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 s

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) HTS for B&W PWRs. Limiting conditions for operation start I' with Section 3 of the TMI-l 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 ?MI-l TS do not define the surveillance requirements based on applicable nodes.

3.1 Startup and Power Operation--Modes 1 and 2 (To erating)

The TMI-l TS require that both reactor coolant loops and both reactor coolant pumps in each loop 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 coolant 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 TMl-1 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, i 3.2 Hot Standby--Mode 3 (T ave >525 F and K eff = 1.00) ,

The TMI-1 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 2

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 ritust 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-l TS do not include the surveillance requirements of MTS 9.4.1.2.1 and 4.4.1.2.2.

3.3 Shutdown--Mode 4 (TAVg525F), Mode 5(TAVG1200 F), K eff10.99 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 T AVG 1 04.

The TMI-l TS do not contain requirements for surveillance of the oper-dbility of the DHR, reactor coolant pumps or steam generator loops in 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 feet 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 specified in MTS 4.9.8.1 and 4.9.8.2.

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4.0 CONCLUSION

S The TMI-l TS do not agree with the MTS in the following areas:

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  • A l (1) They do not follow the MTS format in defining the limiting condi- '

tions by applicable modes, t

(2) They differ from the MTS in Modes 1 and 2 by not requiring reduc-j tion of nuclear power monitor setpoints when operating with less

than four coolant pumps, 4

1 (3) They do not require having at least one coolant l'oop in operation

. in Modes 4 and 5, 1 4

f l (4) They do not require two DHR loops operable in mode 6 with less than 23 feet of water above irradiated fuel assemblies, i

(5) They do not require closing of containment penetrations in Mode 6 with less than one DHR loop operating, j (6) They do not require the surveillance specified in the MTS in l Modes 1 through 6.

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5.0 REFERENCES

l l 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. ,

l i 4. GPU Nuclear letter, H. D. Hukill to NRC, J. F. Stolz, " Decay Heat c Removal (DHR) Technical Specifications," dated January 26, 1982.

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5. Standard Technical Specifications for Babcock and Wilcox Pressurized g ;
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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_ _ __. _ _ _ . . _ _ . _ _ _ . _ _ . . _ . ~ _ . _ . _ . . _ . _ _ . _ _ . _ . . . . . , _ _ _ . . - . ________ _. __ _.._:

1 APPENDIX A M90EL TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT REMOVAL FOR BABC0CK & WILCOX PRESSURIZED WATER REACTORS (PWR's) s)

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r 3/4.4 HEACTOR COOLANT SYSTEM 3/4.4.1 C00LANI LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 4

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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the setpoints for the following trips have been reduced to the values specified in Specification 2.2.1 f or operation with three reactor coolant pumps operating:

1. (Nuclear Overpower).
2. (Nuclear Overpower based on RCS flow and AXI AL 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 iaade while shutdown.

See Special Test Exception 3.10.4.

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REACTOR COOLANT SYSTEM H0f STANDBY LIMITING CONDITION FOR OPERATION

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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 />.
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 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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  • All reactor coolant pumps may be de-energized for up to l 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 100 F below saturation temperature.

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REACTOR COOLANT SYSTEM SHUTDOWN L IMI TING CONDI TION 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 asstciated steam gen-erator and at least one associated reactor coolant pump,
2. Reactor Coolant Loop (8) 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.**

APPLICAdlLITY: MODES 4 and 5.

ACT10N:

a. With less than the above required loops OPERABLE, irrrnediately 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.

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

) (2) core outlet temperature is maintained at least lU FO below saturation temperature.

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REACTOR COOLANT SYSTEM SURVEILL ANCE REQUIREMENT 4.4.1.1.1 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 i 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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REFUEllHG OPERATIONS 3/4.9.8 RESIDUAL Hf AT REMOVAL AND LOOL ANT C IRCULAT10f1 ,

l ALL WATER LEVELS l

.o LIMI TING CONDITI0ff FOR OPERATION b

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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,
n. The DHR loop may be removed from operation for up to l 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 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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REFUFLING OPERATIONS LOW WATER LEVEL LIMITitlG CONDITION FOR OPERATION (.

3.9.8.2 Two independent DHR loops shall be OPERABLE.*

k 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, iinnediately initiate corrective action to return the required loops to l OPERABLE status as soon as possible. l
b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUlk. MENT 4.9.8.2 The required DHR loops shall be determined OPERABLE per Specifica-tion 4.0.5.

l s'C The normal or emergency power source may be inoperable for each DHR loop.

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3/4.4 R E AC TO,( COOLANT SYSTEM BASES 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops in l4 operation, and maintain DNBR above (1.32/1.30) during all normal operations I and anticipated transients. With one reactor coolant pump not in operation l in one loop, THERMAL POWER is restricted by the Nuclear Overpower Based on RCS Flow and AXI AL POWER IMBALANCE and the Nuclear Overpower Based on Pump Monitors trip, ensuring that ti. 'NBR will be maintained above (1.32/1.30) at the maximum possible THERMAL to A ' for the number of reactor coolant pumps in operation or the local qu ' N i at the noint of minimum DNBR equal to (2P/15)%, whichever is more restr.utive.

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

In MODES 4 and 5, a single reactor coolant leap 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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REFUEllf4G OPCRATIOf1S BASES 3/4.9.8 DECAY HEAT REMOVAL Afl0 C00LAf1T CIRCULATI0fl The requirement that at least one DHR loop be in operation ensures A

that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 1400F as required during the REFUEllf4G MODE, and (2) suf ficient coolant circulation is main-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 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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