ML20056H308

Informs That NRR Has Reviewed Proposed Changes & Has No Objection to Proposed Wording Re TS Bases Sections 3/4 3.5.A,3.5.B & 3.5.C

ML20056H308
Person / Time
Site:	Millstone
Issue date:	08/24/1993
From:	James Anderson Office of Nuclear Reactor Regulation
To:	Opeka J CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST NUCLEAR ENERGY CO.
References
TAC-M86940, NUDOCS 9309090135
Download: ML20056H308 (9)

Text

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WASWNGTON, D C. 2055m August 24, 1993 Docket No. 50-245 Mr. John F. Opeka Executive Vice President, Nuclear i

Connecticut Yankee Atomic Power Company Northeast Nuclear Energy Company Post Office Box 270 Hartford, Connecticut 06141-0270 i

SUBJECT:

CHANGE TO TECHNICAL SPECIFICATION BASES SECTIONS 3/4 3.5.%, 3.5.B.

l AND 3.5.C (TAC NO. M86940)

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i By [[letter::05000336/LER-1993-004, :on 930222,reactor Trips on SG Low Water Level Occurred.Caused by Automatic Trip of MFW Pump B on Low Suction Pressure Due to High FW Flow Rates.Classroom Training on Main & Afwc Received by Licensed Operators|letter dated July 2,1993]], Northeast Nuclear Energy Company (NNECO) proposed changes to the Millstone Nuclear Power Station, Unit 1 Technical j

Specification (TS) Bases. NNECO stated that the proposed changes to Bases l

Sections 3.5.A and 3.5.B are necessary as a result of the replacement of the water-cooled low pressure coolant injection (LPCI) pump motors ("C" and "D")

with air-cooled motors qualified to provide long-term post-accident core ana containment cooling. The changes to Bases Section 3.5.C are provided to clarify the basis for feedwater coolant injection (FWCI) subsystem operation, and to identify the components comprising the FWCI subsystem.

Specifically, the changes to Bases Section 3.5.A and 3.5.B remove the limitation of only utilizing the "A" and "B" LPCI pumps for core and containment cooling in the long-term. Since LPCI pumps "C" and "D" are now equipped with air-cooled motors, all the LPCI pumps can be used for core and containment cooling in the long-term and this limitation is no longer neded.

4 The change also deletes the interpretation that the corresponding contair.went cooling subsystem is considered inoperable and in a 4-day limiting condition for operation (LCO) when the "A" or "B" LPCI pump is inoperable. This requirement was in place since the long-term post-accident operation of water-e cooled LPCI pumps could not be assured. Since all four LPCI pumps are now equipped with air-cooled motors, the 4-day LCO for pumps "A" and "B" is no longer needed and the 30-day LC0 stated in the TSs is appropriate.

In l

addition, the staff notes the editorial chtnges to Bases Section 3.5.A and 3.5.B to identify systems consistently and to denote that the limitation on maximum peak clad temperature is less than 2200*F degrees.

j The changes to Bases Section 3.5.C were made to clarify the basis for the FWCI

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subsystem operation, and to properly identify which components comprise the l

FWCI subsystem. The updated basis clarifies that the design basis of the FWCI subsystem is to minimize the probability of the core uncovery, and that the i

FWCI subsystem can provide flow to the core at high pressure in scenarios such as the closure of a main steamline isolation valve or small break loss of coolant accident (LOCA). The changes clarify the redundant emergency core n

cooling systems (i.e., LPCI, core spray, automatic pressure relief, and l

isolation condenser) which are available to mitigate any size LOCA should FWCI I i

fail.

In addition, the staff notes that the statement regarding the assertion i

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t Mr. John F. Opeka August 24, 1993

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the assertion that the FWCI subsystem was inherently more reliable has been removed and that the description of the components which make up a FWCI subsystem have been modified.

The staff has reviewed the proposed changes and has no objection to your proposed wording. Enclosed is a copy of revised TS Bases pages 3/4 5-1, 3/4 5-2, 3/4 5-2a, and 3/4 5-3 for the Millstone Nuclear Power Station, Unit 1.

Sincerely, Original signed by:

James W. Andersen, Acting Project Manager Project Directorate I-4 t

Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

Bases pages cc w/ enclosures:

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Document Name: G:\\ANDERSEN\\86940

Mr. John F. Opeka August 24, 1993 that the FWCI subsystem was inherently more reliable has been removed and that i

the description of the components which make up a FWCI subsystem have been modified.

The staff has reviewed the proposed changes and has no objection to your proposed wording.

Enclosed is a copy of revised TS Bases pages 3/4 5-1, 3/4 5-2, 3/4 5-2a, and 3/4 5-3 for the Millstone Nuclear Power Station, Unit 1.

Sincerely,

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James W. Andersen, Acting Project Manager Project Directorate I-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

Bases pages cc w/ enclosures:

1 See next page f

I l

L Mr. John F. Opeka Millstone Nuclear Power Station Northeast Nuclear Energy Company Unit I L

cc:

Gerald Garfield, Esquire R. M. Kacich, Director Day, Berry and Howard Nuclear Licensing Counselors at Law Northeast Utilities Service Company.

City Place Post Office Box 270 Hartford, Connecticut 06103-3499, Hartford, Connecticut 06141-0270 W. D. Romberg, Vice President J. P. Stetz, Vice President i

Nuclear Operations Services Heddam Neck Plant Northeast Utilities Service Company Ccanecticut Yankee Atomic Power Company Post Office Box 270 362 Injun Hollow Road Hartford, Connecticut 06141-0270 East Hampton, Connecticut 06424-3099 Kevin McCarthy, Director Regional Administrator i

Radiation Control Unit Region I l

Department of Environmental Protection U.S. Nuclear Regulatory Commission 165 Capitol Building 475 Allendale Road Hartford, Connecticut 06106 King of Prussia, Pennsylvania 19406 l

Allan Johanson, Assistant Director First Selectmen Office of Policy and Manacement Town of Waterford Policy Development and Pis:aing Division Hall of Records 80 Washington Street 200 Boston Post Road j

Hartford, Connecticut 06106 Waterford, Connecticut 06385 S. E. Scace, Vice President P. D. Swetland, Resident inspector i

Millstone Nuclear Power Station Millstone Nuclear Power Station Northeast Nuclear Energy Company c/o U.S. Nuclear Regulatory Commission Post Office Box 128 Post Office Box 513 Waterford, Connecticut 06385 Niantic, Connecticut 06357 H. F. Haynes, Nuclear Unit Director G. H. Bouchard, Director Millstone Unit No. I Nuclear Quality Services Northeast Nuclear Energy Company Northeast Utilities Service Company l

Post Office Box 128 Post Office Box 270 l

Waterford, Connecticut 06385 Hartford, Connecticut 06141-0270 Nicholas S. Reynolds Winston & Strawn 1400 L Street, NW Washington, DC 20005-3502 t

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TECHNICAL SPECIFICATION BASES MILLSTONE NUCLEAR POWER STATION. UNIT 1 i

DOCKET NO. 50-245 Replace the following pages c the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

i Remove Insert I

i B 3/4 5-1 B 3/4 5-1 B 3/4 5-2 B 3/4 5-2 B 3/4 5-2a B 3/4 5-2a B 3/4 5-3 8 3/4 5-3 I

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3.5 CORE AND CONTAIMENT COOLING SYSTEMS BASES A.

Core Sorav and LPCI This specification assures that adequate emergency cooling capability is available.

Based on the loss of coolant analysis included in Sections 15.7 and 6.3 of the UFSAR, either of the two core spray subsystems, along with two of the four LPCI pumps, provide sufficient cooling to the core to dissipate the energy associated with the loss of coolant accident and to limit fuel clad temperature (less than 2200*F) to well below the clad melting temperature to assure that core geometry remains intact and to limit any clad metal-water reaction to less than 1%.

In addition, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel.

The accident analysis is additionally conservative in that no credit is taken for spray coolant entering the reactor before the coolant temperature has fallen to 330*F (90 psig).

The LPCI subsystem is designed to provide emergency cooling to the core by ficoding in the event of a loss of coolant accident. This system is completely independent of the core spray subsystem; however, it does function in combination with the core spray system to prevent excessive l

fuel clad temperature. Two pumps of the LPCI subsystem, in combination with a core spray subsystem, provide adequate cooling in the short term

(< 10 min.) for break areas of approximately 0.2 square feet up to and including 4.3 square feet, the latter being the double-ended recirculation line break, without assistance from the high pressure feedwater coolant injection (FWCI) subsystems.

In the long term

(> 10 min.), one of the two LPCI pumps in either loop, in combination 4

I with a core spray subsystem, provide adequate core cooling.

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The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate. The method and concept are described in Reference (1). Using the results developed in this reference, the repair period is found to be less than 1/2 the test interval.

This assumes that the core spray and LPCI subsystems constitute a 1 out of 3 system, however, the combined effect of the two systems to limit excessive clad temperature must also be considered. The test interval specified in Specification 4.5 was 3 months. Therefore, an allowable repair period which maintains the basic risk considering single 1

failures should be less than 45 days and this specification is within this period.

Fcr multiple failures, a shorter interval is specified.

(1) APED 5736, " Guidelines for Determining Safe Test Intervals and Repair Times for Engiraered Safeguards," April 1969, I. M. Jacobs and P. W.

Harriott.

Millstone Unit 1 B 3/4 5-1 Amendment No. J, EE cos!

Ltr. Dtd. August 24. 1993

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3.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES Although it is recognized that the information in Reference (1) provides a quantitative method to estimate allowable repair times, the lack of operating data to support the analytical approach prevents complete acceptance of this method at this time. Therefore, the times stated in the specific items were established with due regard to judgment.

Should one core spray subsystem become inoperable, the remaining core spray and the entire LPCI s.ystem are available should the need for core cooling arise.

Should the loss of one LPCI pump occur, a nearly full complement of core cooling equipment is available. Three LPCI pumps in conjunction with the i core spray subsystem will perform the core cooling function.

Because of the availability of the majority of the core cooling equipment, a 30-day repair period is justified.

If more than one LPCI pump is inoperable.

the repair time is set considering the containment cooling function of the affected LPCI pumps.

B.

Containment Coolina Subsystems The two containment cooling subsystems are provided to remove heat energy from the containment in the event of a loss-of-coolant accident.

Each singlecontainmentcoolingsubsystemincludestwoemeggencyservicewater i

pumps, associated valves, one heat exchanger (40 x 10 BTU /hr), one of the two LPCI pumps and necessary instrumentation, control and power equipment.

It should be noted, the LPCI heat exchanger flow limitation precludes operation of both LPCI pumps in a containment cooling subsystem while performing containment cooling (i.e., with the LPCI heat exchanger i

bypass valve closed). With two heat exchangers (i.e., both loops) operable, it is possible to degrade system performance to one LPCI and one service water pump operating per loop and still not exceed significantly the equipment design temperatures and not rely completely on containment pressure for net positive suction head (NPSH). An interlock to prevent containment spray actuation is included in the j

design of engineered safety features to prevent inadvertent pressure reduction below that required for NPSH. The heat removal capacity of a single cooling loop is adequate to prevent the torus water temperature i

from exceeding the equipment tertperature capability which is specified to be 203*F.

It also provides sufficient subcooling so that adequate NPSH could be assured without reliance on containment pressure except for short intervals during the postulated accident.

In the event that only one heat removal loop is operable, station operation will be permitted for four days unless necessary repairs are made to make the other loop operable. A four-day period is selected to permit reasonable time for operator action and maintenance operations.

i glistoneUnit1 B 3/4 5-2 gndgnt go. k 24,1993 u

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3.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES C.

Feedwater Coolant In_iection Subsystem The Feedwater Coolant Injection (FWCI) subsystem is a single train high pressure Emergency Core Cooling System (ECCS) that utilizes feedwater and condensate components (A or B string) to supply water from the condenser hotwell to the vessel for core cooling. The system consists of the selected motor-driven condensate pump, selected motor-driven condensate booster pump, selected motor-driven feedwater pump, and associated piping and valves.

In addition, the system also includes a selector switch for selecting the desired FWCI pumps (A or B) and the emergency condensate i

transfer (ECT) pump to supply make-up water to the condenser hotwell from the Condensate Storage Tank (CST).

i The FWCI subsystem is automatically initiated on an ECCS actuation signal (high DW Pressure or low-low RPV level). Upon receipt of an initiation i

signal, if off-site power remains available, the feedwater system will continue to run either in the level control mode or in the flow control mode until the train trips when the condenser hotwell water is depleted.

In the event of loss of off-site power coincident with an initiation signal, power to the preselected feedwater string (A or B) and ECT pump are supplied by the Gas Turbine (GT) Generator. Regardless of availability of off-site power, the ECT pump is automatically started upon receipt of an initiation signal, and will provide make-up water to the condenser hotwell from the CST.

The design function of the FVCI subsystem is to reduce the probability of core uncovery during the-lifetime of the plant. The FWCI subsystem can perform this function by replenishing reactor inventory during postulated high pressure events (e.g., MSIV closure) thereby avoiding reactor blewdown and potential partial core uncovery. The FWCI subsystem can also provide core cooling for small break LOCA while the RPV is still pressurized. However, redundant ECCS systems (i.e., LPCI, CS, APR and IC) are available to mitigate any size LOCA should the FWCI subsystem fail.

When the FWCI subsystem is powered by the GT Generator, its flowrate is automatically reduced to ensure that the GT is not being overloaded.

Also, the 14A to I4G tie breaker trip protects the GT from being overloaded.

Millstone Unit I B 3/4 5-2a Amendment No.

ess Ltr. Dtd. August 24, 1993

3.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES D.

Automatic Pressure Relief (APR) Systems The relief valves of the automatic pressure relief subsystem are a backup i

to the FWCI subsystem. They enable the core spray or LPCI to provide protection against the small pipe break in the event of FWCI failure, by depressurizing the reactor vessel rapidly enough to actuate the core sprays or LPCI. The core spray and/or LPCI provide sufficient flow of coolant to limit fuel clad temperature to well below clad melt to assure that core geometry remains intact.

^

A manual actuation of each APR valve is performed to verify the valve and solenoid are functioning properly and no blockage exists in the SRV discharge lines.

It has been demonstrated that the blowdown of an APR/SRV to the torus causes a wave action that is detectable on the torus water level instrumentation. The discharge of a relief line is audible to an individual located outside the torus in the vicinity of the line, as experienced at other BWRs.

Adequate reactor steam dome pressure must be available to perfom this test to avoid damaging the valve. This test can be performed prior to the mode switch being placed in the RUN position but in no case more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the switch being placed in the RUN position. This 1

requirement is sufficient to allow testing at high enough pressure, while limiting reactor power to less than 25 percent power ensures that the 1

remaining APR/SRV are adequate to handle any transient in the unlikely event that one ARP/SRV discharge line is found blocked. Reactor start-up is allowed prior to performing this test because valve operability and the set points for overpressure protection are verified, per ASME requirements, prior to valve installation.

E.

Isolation Condenser Svstem The turbine main condenser is normally available. The isolation condenser is provided for core decay heat removal following reactor isolation and scram. The isolation condenser has a heat removal capacity sufficient to handle the decay heat production at 300 seconds following a scram. Water will be lost from the reactor vessel through the relief valves in the 300 seconds following isolation and scram. This represents i

a minor loss relative to the vessel inventory.

I g11stoneUnit1 B 3/4 5-3 gmendgegt fug'uk 24,1993

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