ML20244D889

Submits Info Re Station Blackout Rule 10CFR50.63,including Justification for Selection of Proposed Blackout Duration, Description of Procedures Implemented for Events & Discussion of Plant Unique Design Re Survival Capability

ML20244D889
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	04/17/1989
From:	Berry K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 8904240181
Download: ML20244D889 (6)

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s' Consumers POVrar Kenneth W Berry MMY Director MMM M Nuclear licensing General Offu:es: 1945 West Pernall Road, Jackson. MI 49201 e (517) 788-1636 April 17, 1989 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-155 LICENSE DPR BIG ROCK POINT PLANT -

RESPONSE TO STATION BLACKOUT RULE 10 CFR 50.63 On July 21, 1988, the Nuclear Regulatory Commission (NRC) amended its regulation in 10 CFR, Part 50. A new section, 50.63, was added which requires that each light-water-cooled nuclear power plant be able to withstand and recover from a station blackout (SBO) of a specified duration. Utilities are expected.to have the baseline assumptions, analyses and related information used in their coping evaluation available for NRC review. It also identifies the factors that must be considered in specifying the station blackout duration. Section 50.63 requires that, for the station blackout duration, the plant be capable of maintaining core cooling and containment integrity.

Section 50.63 further requires that each licensee submit the following information:

1. A proposed station blackout duration including a justification for the selection based on the redundancy and reliability of the onsite emergency ac power sources, the expected frequency of loss of offsite power, and the probable time needed to restore offsite power;

2. A description of the procedures that will be implemented for station blackout events for the duration (as determined in 1 above) and for recovery therefrom; and

3. A list and proposed schedule for any needed modifications to equipment and associated procedures necessary for the specified SB0 duration.

The NRC has issued Regulatory Guide 1.155 " Station Blackout" which describes a means acceptable to the NRC Staff for meeting the requirements of 10 CFR 50.63. Regulatory Guide (RG) 1.155 states that the NRC Staff has determined that NUMARC 87-00 " Guideline and Technical Bases for NUMARC P

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Big Rock Point Plant Station Blackout Rule 10CFR50.63 April 17, 1989 Initiatives Addressing Station Blackout At Light Water Reactors" also provides guidance that is in large part identical to the RG 1.155 guidance and is acceptable to the NRC Staff for meeting these requirements.

Table 1 of RG 1.155 provides a cross-reference between RG 1.155 and NUMARC 87-00 and notes where the RG takes precedence.

Consumers Power Company has evaluated the Big Rock Point Nuclear Power Plant against the requirements of the SB0 rule using guidance from NUMARC 87-00 except where RG 1.155 takes precedence. The results of this evaluation are described below. (Applicable NUMARC 87-00 sections are shown in parenthesis). All sections of this generic response are identical to the recommended NUMARC generic response except where indicated with an nen, 1

A. Proposed Station Blackout Duration l NUMARC 87-00, Section 3 was used to determine a proposed SB0 duracion of four hours.

The following plant factors were identified in determining the proposed station blackout duration:

1. AC Power Design Characteristic Group is P3 based on:

a. Expected frequency of grid-related LOOPS - does not exceed once per 20 years (Section 3.2.1, Part 1A p. 3-3);

• Actual data supporting this conclusion was unavailable. NUREG-1032 notes sites having a frequency of grid-related events at the once per 20 site-year frequency are limited to St. Lucie, Turkey Point and Indian Point. Accordingly, no other sites are expected to exceed the once per 20 site-year frequency of grid-related loss of off-site power events.

b. Estimated frequency of LOOPS due to extremely severe weather places the plant in ESW Group 2 (Section 3.2.1, Part IB, p. 3-4); ,

c. Estimated frequency of LOOPS due to severe weather places the plant in SW Group 5 (Section 3.2.1, Part IC, p. 3-7);

d. The offsite power system is in the I2 Group (Section 3.2.1, Part ID, p. 3-10);

2. The emergency AC power configuration group is A based on: (Section 3.2.2, Part 2C, p. 3-13)

a. There is one emergency AC power supply not credited as alternate AC power sources (Section 3.2.2, Part 2A, p. 3-15);

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Station Blackout Rule 10CFR50.63 April 17 -1989

b. No emergency AC power supplies'are necessary to operate safe shutdown equipment following a loss of.off site power (Section 3.2.2,-Part 2B, p. 3-15).

• Because of Big Rock Point's unique capability to cope with a loss of offsite power without ac power, an EAC. Power Configuration group was impossible.to determine directly from Table 3, Reg. Guide 1.155. An assumption has been made that Big Rock Point's EAC configuration group is A, based on the following; The EAC configuration groups are based on the redundancy of EAC.

power systems, the greater the EAC system redundancy, the.less of a potential-for a station blackout. Group A is characterized by highly redundant and independent EAC sources. Big Rock Point's lack of need for AC power is similar-to a plant whose EAC power sources are highly redundant and independent, both set of circumstances drastically reduce the potential of a station blackout.- ,

3. The target EDG reliability is 0.975.

a. A target EDG reliability of 0.975 was selected based on having a nuclear unit average EDG reliability for the last 100 demands greater than 0.95; B. Procedere Description Plant procedures have been reviewed and changes necessary to meet the guidelines in NUMARC 87-00, Section 4 will be implemented in the'following areas.

1. Station Blackout response procedures; ONP 2.36 Loss of Station Power

2. AC power restoration procedures; Offsite Power Restoration

3. Severe weather procedures;

EMP 3-7 Earthquakes <

EMP 3-8 Severe Weather l 1 '

• C. Coping Analysis A coping analysis was not performed to determine if Big Rock Point could survive a loss a off-site power for the determined 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> duration. Big Rock Point is capable of surviving a loss of off-site power event without the use of AC power for an indefinite period of time due to its unique design features including only one automatic diesel generator and requiring no diesel generator for decay heat removal. A brief discussion of Big Rock  !

Point's unique design will clarify the issue; Big Rock Point has an emergency condenser (EC) (similar to the isolation condenser at newer plants) located within the containment building, capable of removing 100% of the decay heat generated following a SCRAM. Actuation l of the emergency condenser is automatic. Upon total loss of station power, OC0489-003S-NL01 L_________________________- 1

Nuclear Regulatory Commission 4 Big Rock Point Plant Station Blackout Rule 10CFR50.63 April 17, 1989 the Reactor Protection System (RPS) generates an isolation scram signal.

The DC powered Main Steam Isolation Valve (MSIV) closes and the de powered emergency condenser outlet valves (MO-7053 and M0-7063) automatically open.

Cooling water make-up to the shell side of the emergency condenser is not necessary for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following actuation. At that time cooling water can be supplied by the diesel driven fire pump via a de solenoid operated make-up valve if ac power is not yet available. Shell side low level annunciation is available in the control room and the alternate shutdown building. Reactor pressure indication is also available in the alternate shutdown building.

The emergency condenser depressurizes the primary system down to 140 psi in 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, assuming that continued make-up to the shell side of the emergency condenser is available and a 100 degree per hour cooldown rate as specified in ONP-2.36. Once primary system pressure is reduced, primary system make-up can be supplied by the diesel driven fire pump threugh the de operated core spray valves (MO-7051 and M0-7061). Make-up to the primary system is not necessary for 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br />, at which time, the reactor low water set point is reached, assuming a leak rate of 10 gpm (technical specifica-tion limitation).

D. Proposed Modifications and Schedule

• The ability of Big Rock Point to cope with a station blackout was assessed with the following results:

1) Decay Heat Removal It has been determined that decay heat is sufficiently removed from the primary system following a loss of str. tion power for an indefinite period of time without ac power using the emergency condenser.

Circulation from the PCS to the emergency condenser proceeds by natural circulation. No active transport is required. No plant modifications and/or procedure changes are necessary.

2) Emergency Condenser Make-up It has been determined that shell side make-up can be supplied from the diesel driven fire pump without an ac power dependence. The water supply is from Lake Michigan and is considered adequate. No plant modifications and/or procedure changes are necessary.

3) Reactor Coolant Inventory It has been determined that Reactor Coolant Inventory make-up is not necessary until after the pressure has decreased below the shut off head of the fire pump. At this time, the diesel driven fire pump can supply make-up to the primary system without ac power. No plant modifications and/or procedure changes are necessary.

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4) Alternate Shutdown Building (ASDB) Battery Capacity It has been determined that the ASDB battery capacity is sufficient to operate the MSIV the Emergency Condenser Outlet valves and the solenoid operated nakeup valve. No plant modifications and/or procedure changes  !

are necessary.

5) Station Battery Capacity The station blackout coping analysis performed as a part of 10CFR50 Appendix R modifications indicate that the plant can be maintained in the hot shutdown' condition without ac power for at least one week before required battery supplies are depleted below acceptable voltage.

Following battery depletion, manual operation of the core spray valves permit maintenance of the hot shutdown condition indefinitely. No plant modifications and/or procedure change are necessary.

6) Compressor Air All air operated valves necessary to cope with a SB0 are of the fail safe nature. No plant modifications and/or procedure changes are necessary.

7) Effects of Loss of Ventilation The following areas have been identified as areas of concern:

AREA EQUIPMENT ASDB Power supply for MSIV, SV-4947 and Emergency Condenser valves.

Reactor pressure indication Screenhouse Diesel Fire Pump Electrical Power supply for MO-7051 and Equipment Room )

MO-7061. I

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Control Room M0-7051 and M0-7061 controls, j l

Containment Emergency Condenser and valves and MSIV. J Reasonable assurance of the operability of station blackout response equipment in the above areas of concern has been assessed. No modifications or associated procedure changes are required to provide reasonable assurance for equipment operability.

8) Containment Isolation The plant list of containment isolation valves has been reviewed to verify that valves which must be capable of being closed or that must be operated (cycled) under station blackout conditions can be l

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positioned (with indication) independent of the preferred and blacked-out unit's Class 1E power supplies. No plant modifications and/or q associated procedure changes were determined to be required to ensure that appropriate containment integrity can be provided under SB0 l 1

conditions.

9) EDG Reliability Monitoring Program I l

An EDC Reliability Monitoring Program will be implemented. Imple-mentation schedules and status reports will be included in the Integrated Plan Updates.

Procedure changes identified it. Part B above will be completed by January 1990 in accordance with 10 CFR 50.63(c)(3).

f f W0<wf Kenneth W Berry Director, Nuclear Licensing CC Administrator, Region III, USNRC NRC Resident Inspector - Big Rock Point NUMARC 1

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