ML20044A507

Forwards Revised Response to Station Blackout Rule for Plant.During Blackout Event,Plant Can Utilize RCIC Sys or HPCS to Provide Required Reactor Vessel Inventory Makeup

ML20044A507
Person / Time
Site:	LaSalle
Issue date:	06/22/1990
From:	Richter M COMMONWEALTH EDISON CO.
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9006290136
Download: ML20044A507 (11)

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1. . Commonwealth Edison 1400 Opus Place

/ Downers Crove. Illinois 60616 s

0$4 .'2, 1990 Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20SS5 Attn: Document Control Desk

Subject:

LaSalle County Station Units 1 and 2 Revised Response to Station Blackout Rule NRC_DochtLNos. 50-373 and_5h374

References:

(a) 10 CFR Part 50.63, Loss of all Alternating i Current Power.  !

(b) H. Richter (Ceco) letter to T.E. Hurley (NRC),

dated April 17, 1989.

(c) B. Lee (NUMARC) letter to NUMARC Board of Directors, dated January 4, 1990.

(d) H. Richter (CECO) letter to T.E. Hurley (NRC),

dated March 30, 1990.

Dr. Hurley:

Reference (a) requires that each light-water-cooled nuclear power plant be able to withstand and recover from a station blackout (SBO) of a specified duration. A response to the SB0 rule was required from each Ilcensee by April 17, 1989. Reference (b) provided Commonwealth Edison Company's (CECO) initial response to the SB0 rule for LaSalle County Station.

Subsequent to the initial response, Reference (c) provided additional guidance / clarifications to the NUMARC 87-00 guidelines. As a result of-these NUMARC clarifications and CECO /NRC staff discussions on the SB0 rule, Ceco has re-evalt.ated the original response for LaSalle County Station and determined that a revised response was necessary.

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Dr. Thomas E. Hurley June 22, 1990 '

Attachment 'A' to this letter provides CECO's revised response to the SB0 rule for laSalle County Station. Ceco is-proposing to meet the SB0 rule for LaSalle County Station by demonstrating that the station can cope for the four-hour blackout duration. During a blackout event, LaSalle County Station can utilize the Reactor Core Isolation Cooling (RCIC) System or the High  ;

Pressure Core Spray (HPCS) System to provide the required reactor vessel inventory makeup. An Alternate AC power source exists for each unit which >

meets the criteria specified in Appendix B of NUMARC 87-00, and provides the  :

necessary power for HPCS System availability.

Please address any questions that you may have concerning this  !

response to this office.

Respectfully, f .4. ,

M.H. Richter Generic Issues Administrator Attachment 'A': Response to Station Blackout Rule for LaSalle County Station. !

cc: A.B. Davis, Regional Administrator - RIII Senior Resident Inspector - LaSalle Station i R. Pulsifer - NRR Project Manager P. Gill - NRR Electrical Systems Branch i

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ATTACHMENT A RESPONSE 10 STATION BLACK 00i ROLE FOR LASALLE COUNTY STATION M

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RERONSLJQ_SIAHON_RL6CK0MLRulE LOR _LASALLLCOUNILSlAIIDE On July 21, 1988, the Nuclear Regulatory Commission (NRC) amended its regulations 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 w'thstand and recover from a station blackout (580) of a specified duration. Licensees were required to submit station blackout responses by April 17, 1989. In accordance with the rule. Commonwealth Edison Company (CECO) submitted a response for LaSalle County Station on that date.

Following the NRC review of several licensee responses, it was determined that additional guidance / clarifications were necessary for the industry. The NRC and NUMARC subsequently agreed upon a set of clarifications to the NUMARC guidance consisting of a set of questions and answers. This supplement to NUMARC 87-00 was issued to the industry by NUMARC on January 4, 1990.

Prior to the issuance of the NUMARC letter (January 4, 1990), CECO had been involved in discussions with the NRC staff on the SB0 responses for Dresden and Quad Cities Stations. As a result of the NUMARC clarifications and the Ceco /NRC staff discussions, CECO re-evaluated the original _LaSalle Station response and coping calculations, and determined that a revised response was necessary, for this response. CECO has evaluated LaSalle Station against the requirements of the SB0 rule using the recent clarifications of NUMARC 87-00 except where Regulatory Guide 1.155 takes precedence. In some instances, alternative technical methodologies were used. These alternate approaches are highlighted in this response where appropriate.

This response details the plant factors considered in the determination of the proposed station blackout coping duration. In addition, the ability of LaSalle Station to cope with a station blackout of this proposed duration is addressed. Finally, procedure revisions and modifications required to conform to the guidance are described. The results of this evaluation are detailed in this response. (Applicable NUMARC 87-00 sections are shown in parentheses.)

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A. Et0001edJtation_ Bitch 0ULDunt100 NUMARC 87-00, Section 3 combined with the recently issued guidance, was used to determine a proposed SB0 duration of four hours. The plant factors considered in determining the proposed SB0 duration are discussed in this section. Additionally, the available Alternate AC power source option is discussed. A simpilfled one-line diagram of LaSalle Station's AC electrical distributton is presented on Figure 1. ,

1. The AC Power Design Characteristic Group is P1 based on: l

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

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

c. Estimated frequency of LOOPS due to severe weather places the  ;

plant in SW Group 2 (Section 3.2.1 Part IC, p. 3-7),

d. The off-site power system is in the I 1/2 Group (Section 3.2.1, Part ID, p. 3-10). i

2. The Emergency AC power configuration group is D (Section 3.2.2, Part 2C, L. 3-13) based on:

a. There are three emergency AC power supplies not credited as Alternate AC power sources (Section 3.2.2, Part 2A, p. 3-15); >

b. Two emergency AC power supplies are necessary to operate safe shutdown equipment for an extended period following a loss of off-site power (Sectlen 3.2.2, Part 2B, p. 3-15).

3. The target EDG reliability to be maintained by the site 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, consistent with NUMARC 87-00, Section 3.2.4. ,

b. A diesel generator reliability program incorporating the five elements discussed in Regulatory Guide 1.155 will_ be established to ensure this target is maintained. In addition, Ceco is monitoring the resolution of Generic Issue B-56: Diesel Generator Reliability. When the final guidance on the resolution of this issue is published, CECO will review, and if necessary, revise the program.

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4. An Alternate AC (AAC) power source option exists at LaSalle Station which meets the criteria specified in Appendix B to NUMARC B7-00.

An AAC power source exists for each unit, and is available within ten .

minutes of the onset of the station blackout event. The AAC power source has sufficient capacity to operate systems capable of coping with a station blackout for the required 500 duration of four hours.

The AAC power source for each unit is the Division 3 emergency diesel generator (D/G 18 for Unit I and D/G 2B for Unit 2) which supplies power to the HpCS System and its auxillaries. The AAC division (Division 3) includes its own safe shutdown equipment which is physically and electrically isolated from the unit's normal safe shutdown busses (busses 141-Y and 142-Y for Unit I, and busses 241-Y ,

and 242-Y for Unit 2). The HPCS System is designed to provide  :

coolant injection at all reactor pressure conditions, and is casable of maintaining the plant in hot shutdown for the expected four-lour duration of the station blackout, i

This AAC power source is not susceptible to a single point  ;

vulnerability whereby a likely weather-related event or single active failure could disable any portion of the emergency ac power sources  ;

or the preferred power supply, and simultaneously fall the AAC source. Since this AAC power system 15 designed to power safe shutdown loads during a loss of offsite power event, a transient loading analysts was not necessary.

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B. EtopostLConinLasitumeAt The ability of LaSalle Station to cope with a four-hour station blackout l In accordance with NVKARC 87-00 Section 3.2.5 is assessed in this section using NUMARC 87-00, Section 7. This coping assessment considers (1) the adequacy of the condensate inventory, (2) the capacity of the Class IE l batteries, (3) the station blackout compressed air requirements, (4) the i effects of loss of ventilation on station blackout response equipment, and  !

(5) the ability to maintain containment integrity. The coping assessment l calculations referenced in this section are available for review.

1. CODdensiLteJAYentory. lor._DenLHeillemonLiittilon 7.2.1)/Reaciqr Cooj apt _ Inventory _lSetiloLL51 ,

It has been conservatively determined from Section 7.2.1 of NUHARC 87-00 that 167,000 gallons of water are adequate for decay heat removal for f our hours. The station blackout coping approach uses the Reactor Core Isolation Cooling (RCIC) System or the HPCS System to provide makeup water for core cooling. The HPCS system normally takes suction from the suppression pool which contains approximately 950,000 gallons of water. The RCIC System normally. takes suction  !

from the condensate storage tank (CST) and automatically transfers to the suppression pool on low CST level. Since decay heat is removed by the discharge of steam through the main steamline safety / relief valves (SRVs) into the suppression pool, this source will not be significantly depleted by RCIC/HPCS operation. A plant-specific analysis was performed to ensure that the RCIC System or HPCS System

, can supply sufficient condensate to maintain the core covered for the -

l entire four-hour duration of the station blackout. ,

As a result of SRV discharge, gradual heatup of the suppression pool will occur. A calculation was performed to ensure that the suppression pool temperature would not exceed the temperature / pressure limits within four hours. A leakage rate of 18 gpm from each reactor recirculation pump was assumed in this

! analysis. In addition, the Technical Specification leakage rate of 25 gpm (20 gpm identified plus 5 gpm unidentifled) was included.

No plant modifications or procedure changes are needed to uttilze the water sources (CST and suppression pool).

2. ClutlLDAtt ery_CapacliLIS.e.c t i on 7 . 2 . 2 )

The 125 Vdc (Divisions 1 and 2) and 250 Vdc Class 1E batteries are being replaced with larger capacity batteries. At this time, the 125 l

Vdc Division 1 batteries for both units, and the 250 Vdc battery for i Unit 2, have been replaced. A calculation was performed to ensure ,

that these new batteries have sufficient capacity to meet the station '

blackout loads for four hours assuming that loads not needed to cope with a station blackout are shed. The required loads include power restoration from either the emergency ac power supplies or the preferred power source. The loads that need to be shed are listed in the battery capacity calculation performed for the SB0 analysis. The shedding of these loads will be proceduralized as noted in Section C' of this response.

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3. CompIt11tL61L.(S.tttion 7.2.3)

The AAC power source does not power station t.ompressors, however, instrument nitrogen is required for the relief mode operation of the main steamline SRVs. The Automatic Depressurization System (ADS) valvet (7 of the SRV's) have existing backup nitrogen bottle banks that have been analyzed to ensure they are sufficient to support SRV.

actuations for the four-hour coping duration. Manual opening and closing of individual ADS valves, to depressurize the _ reactor in a controlled manner, requires sending an operator to the Auxiliary Electric Equipment Room (AEER). Emergency lighting and communications already exist to gain access to the AEER and to utilize the required controls. Control of the ADS valves from the AEER can be established within 20 minutes. During this time, the Low Low Set function of five of the ADS valves will automatically control reactor pressure in the normal operating range. The ADS valves (all 7 at once) can be manually initiated by either of two divisions of DC logic from the Control Room. As a backup, the mechanical safety node of SRV operation is available independent of nitrogen bottles to control pressure.

4. Iffnis_oL_Lossaf Ventilation _iSection 7.2.4_)

The AAC power source does not power cooling loads to some plant areas that contain station blackout response. equipment. These areas must therefore be evaluated to establish reasonable assurance of operability for station blackout equipment. This section documents that reasonable assurance of operability is provided for the containment and all dominant areas of concern.

In addition, the Residual Heat Removal (RHR) System is not energized during a station blackout. Therefore, a calculation has been i performed to ensure that the suppression pool temperature / pressure limit will not be violated under station blackout conditions,

a. Doninni_Attis_of_ Concern The dominant areas of concern (DACs) at LaSalle Station were chosen from rooms that, based on documented engineering judgment, (1) contained station blackout response equipment,- (2) have substantial heat generation terms, and (3) lack normal heat removal systems due to the blackout. These areas are-listed in the following table along with their associated station blackout temperature, type of heatup analysis performed, and justification for Reasonable Assurance of Operability (RAO). 1 AREA EQUR HR. TEMP.- AML151S M0llSIlflCAIL0li Aux Elect Equip Rooms 117'F transient less than 120'F (non-NUMARC)

Control Room 98.1'F transient- less than 120*F (non-NUMARC)

RCIC Room 153'F transient NUMARC 87-00 (non-NUMARC) App. F Report (A-5)

Reasonable assurance of equipment operability is established '

- without further analysts if temperatures in the DAC are calculated to be equal to or less than 120'f (NUMARC 87-00 Supplemental Questions / Answer #2.2). Procedure revisions are required for opening access and panel doors in the Aux 111ary  :

Electrical Equipment Rooms. No modifications are required to provide reasonable assurance of equipment operability in the above rooms.

b. Contajnment i The AAC power supply does not energize containment cooling ,

systems. Therefore, a loss of ventilation analysis was performed '

for the containment under station blackout conditions. This analysis verifles the NUMARC 87-00 assumption (Section 2.7.1) that containment temperatures resulting from a station blackout are enveloped by those of the loss of coolant accident.

c. Suppression _ fool The AAC power supply does not energize the RHR System. ,

Therefore, the suppression pool temperature will increase due to the discharge of steam from the safety / relief valves. A calculation was performed which ensured that the suppression pool temperature would not exceed the temperature / pressure limits within four hours. A leakage rate of 18 gom from each reactor recirculation pump was assumed in this analysis. In addition, the Technical Specification leakage rate of 25 gpm (20 gpm #

identified plus 5 gpm unidentified) was included.

5. Con t atamenLis olation_1S e c tioILLL.51 The station list of containment isolation valves has been reviewed to verify that valves which must be capabic of being closed, or that ,

must be operated (cycled) under station blackout conditions, can be positioned (wlth indication) independent of the preferred or Class IE <

power supplies. This analysis used the exclusion criteria listed in NUMARC 87-00, Section 7.2.5. When multiple valves are provided in a line penetrating containment, only one valve is required to be closed.

The valves that may require manual actuation to ensure appropriate containment integrity under station blackout conditions will be ,

incorporated into the appropriate station procedures. The associated procedure revisions are documented in Section C of this response.

6. Qualitylssurance A QA program meeting the requirements of Regulatory Guide 1.155-Appendices A and B will be applied to cover non-safety related equipment needed for coping with a station blackout that are not already covered by existing QA requirements in Appendices B or'R of 10 CfR 50.

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. C. Propo s e d _EE0c thttf Lauft hd1 LLC a dDn l This section documents the proposed procedure revisions and modifications for LaSalle Station to conform to the station blackout rule, I. ELOCtdULtle.YhtDR$

The following potential procedure revisions have been identified to meet the station blackout rule.

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10PIC ._fROCiDURL#_ NAIURLOLREVISION.__

Battery Capacity LOA-AP-08 shed appropriate battery loads Loss of Ventilation LOA-AP-08 open AEER access and panel doors Containment Isolation LOA-AP-98 valves requiring manual actuation ,

Containment Isolation LOA-PC-01 valves requiring manual actuation [

Containment Isolation LOA-HS-02 valves requiring manual actuation Containment Isolation LOA-RH-04 valves requiring manual actuation Containment Isolation LOA-RI-03 valves requiring manual actuation Containment Isolation LOA-RT-02 valves requiring manual actuation Containment Isolation LOS-HG-SAI valves requiring manual actuation ,

Restoration of AC power SpS0-1-1 system load dispatcher guidanco The L5CS housekeeping program includes periodic inspections of the site with the Intent of minimizing the accumulation of excessive debris, This '

satisfies the intent of the severe weather actions per NUHARC 87-00 Section 4.2.3.

l These procedure revisions will be completed one year-after the- ,

notification provided by the Director, Office of Nuclear Reactor Regulation in accordance with 10 CFR 50.63 (c)(3).

2. Hod.llications The 125 Vdc (Divisions I and 2) and 250 Vdc Class IE batteries are being .

replaced with larger capacity batteries. These modifications are scheduled to be completed by the end of the fourth refueling outage for each unit (Unit 1 - scheduled for April 1991, Unit 2 - scheduled for l January 1992).

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