View source for B13179, Responds to NRC Station Blackout Rule Re Capability of Plant to Maintain Core Cooling & Containment Integrity. Procedure Changes Identified in Part B of Attachment 1 Will Be Completed in 1 Yr After Notification Provided by NRR

{{Adams
| number = ML20244D566
| issue date = 04/17/1989
| title = Responds to NRC Station Blackout Rule Re Capability of Plant to Maintain Core Cooling & Containment Integrity. Procedure Changes Identified in Part B of Attachment 1 Will Be Completed in 1 Yr After Notification Provided by NRR
| author name = Mroczka E
| author affiliation = CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST UTILITIES
| addressee name = Murley T
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| docket = 05000213
| license number = 
| contact person = 
| document report number = B13179, NUDOCS 8904240005
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 8
}}

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April 17, 1989                                            l Docket No. 50-213 B13179 Re:    10CFR50.63 ISAP Topic 1.116 Dr. T. E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

==Dear Dr. Murley:==

I Haddam Neck Plant                                                       ;
Resoonse to Station Blackout Rule On June 21, 1988,(I) the Nuclear Regulatory Commission (NRC) amended its regulations in 10CFR50 (effective date July 21, 1988). 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 (SB0) of a specified                                           ,
duration.      Licensees are expected to have the baseline assumptions, analyses,                                     i and related information used in their coping evaluation available for NRC Staff review. It also identifies the factors that must be considered in specifying the SB0 duration. Section 50.63 requires that for the SB0 dura-tion. the plant be capable of maintaining core cooling and :cpropriate con-tainment integrity. Section 50.63 further requires that each licensee submit the following information:
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: 1. A proposed SB0 duration including a justification for the selection based on the redundancy and reliability of the on-site emergency AC power sources, the expected frequency of loss of off-site power, and the probable time needed to restore off-site power.
: 2. A description of the procedures that will be implemented for SB0 events for the duration (as determined in 1 abe"c) r:f f:r r:::c:ry ther:fr=.
: 3. A list and proposed schedule for any needed modifications to equipment                                          !
l                               and associated procedures necessary for the specified SB0 duration. ~
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l (1) 53 Federal Register 23203, June 21, 1988.
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l                                    8904240005 890417 PDR       ADOCK 05000213 P                             PNV

     . Dr. T. E. Murley B13179/Page 2 April 17, 1989 The NRC Staff has issued Regulatory Guide 1.155, " Station Blackout," which describes a means acceptable to the NRC Staff for meeting the requirements of 10CFR50.63. Regulatory Guide (RG) 1.155 states that the NRC Staff has deter-mined that Nuclear Management and Resources Council (NUMARC) 87-00 " Guidelines and Technical Bases for NUMARC Initiatives Address 1hg 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.
Connecticut Yankee Atomic Power Company (CYAPC0) has evaluated the Haddam Neck Pl ant against the requirements of the SB0 rulf                                                              guidance from NUMARC 87-00 except where RG 1.155 takes precedence."g       CYAPC0 using                                              has determined that the Haddam Neck Plant is in the 4-hour required coping duration category and that the plant has satisfactorily assessed its ability to cope with an SB0 for this time interval using Section 7 of NUMARC 87-00, The details of this evaluation are nrovided in Attachment No.1.
Attachment No.1 follows the standard SB0 rule gsponse format developed by NUMARC and formally approved by the NRC Staff            as providing an adequate level of information for the review.        All documentation of the calculations and procedure reviews is maintained in our files and is available for NRC Staff review.
The NRC Staff states that a goal of the SB0 rule is to maintain the frequency of core damage from SB0 near or below 10 5 per reactor-year. In the supple-mentary information of the final rule, the NRC Staff states that the SB0 rule must be met regardless of whether a plant-specific probabilistic risk assess-ment (PRA) currently meets this goal. The NRC Staff does not, on the other hand, preclude the licensee from identifying plant-specific PRA data to support a determination that SB0 would have an acceptably small probability forcagngcoredamage. Accordingly, CYAPC0 reiterates our previous determi-nation      that the core melt frequency of an SB0 at the Haddam Neck Plant from internally initiated events at power is approximately 2 x 10 5 per reactor-year (i.e., exclusive of external events as defined in the PRA Procedure Guide NUREG/CR-2300). Therefore, this information, along with the results of the NUMARC 87-00 evaluation, substantiates that the risk of core alt from an SB0 at the Haddam Neck Plant is acceptably low.
(2) Table 1 to RG 1.155 provides a cross-reference between RG 1.155 and NUMARC 87-00 and notes where the RG takes precedence.
(3)   A. C. Thadani letter to W. H. Rasin, dated October 7, 1988.
(4)   J. F. Opeka letter to C. I. Grimes, "ProbaL''istic Safety Study--Summary Report and Results," dated March 31, 1986.

       . Dr.: T. E. Murley B13179/Page 3 April 17, 1989 An Integrated Safety Assessment Program (ISAP) Topic No.1.116, " Station
           ~ Blackout," was initiated to evaluate specific plant modifications which resulted from the evaluation conducg in response to the NRC rulemaking and industry initiatives on this issue.         Since this evaluation concludes that
;            the risk of core melt from SB0 is acceptably low,. as detailed in Attach-ment No.1, no further evaluations in ISAP are necessary.          CYAPC0 plans to close ISAP Topic No.1.116, " Station Blackout," pending receipt of the NRC Staff response to this transmittal and subsequent implementation of applicable                      .
procedure changes.                                                                                  l The procedure changes identified in Part B of Attachment No. I will be com-pleted within one year after notification is provided by the Director, Office                       1 of Nuclear Reactor Regulation, in accordance with 10CFR50.63(c)(3).
If you have any questions, please contact us.
Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY 4H E.J.Mroczkaf Sen or Vice President cc:   W. T. Russell, Region I Administrator A. B. Wang, NRC Project Manager, Haddam Neck Plant J. T. Shedlosky, Senior Resident Inspector, Haddam Neck Plant U.S. Nuclear Regulatory Commission Attention: Document Control Desk
;                  Washi.igton, DC 20555 NUMARC 1776 Eye Street, NW Suite 300 Washington, DC 20006-2496 (5)   E. J. Mroczka letter to    U.S. Nuclear Regulatory Commission Document Control Desk, " Integrated Safety Assessment Program," dated March 2, 1989.
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Docket No. 50-213 B13179 ISAP Topic 1.116 Attachment No. 1 Haddam Neck Plant Response to SB0 Rule l
April 1989
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1 Dr. T. E. Murley i
                   *B13179/ Attachment No. 1/Page 1 April 17, 1989 Haddam Neck Plant Response to SB0 Rule A. Proposed SB0 Duration NUMARC 87-00, Section 3 was used to determine an SB0 coping duration category of 4 hours.       No modifications were required to attain this proposed coping duration category.
The following plant factors were identified in determining the proposed SB0 coping duration:
: 1. AC Power Design Characteristic Group is P1 based on:
: a. Expected frequency of grid-related loss of off-site power (LOOP) does not exceed once per 20 years (NUMARC 87-00 Sec-tion 3.2.1, Part 1A, page 3-3).
: b. Estimated frequency of LOOP due to egemely severe weather (ESW) places the plant in ESW Group 1         (NUMARC 87-00, Sec-tion 3.2.1, Part IB, page 3-4).
: c. Estimated frequency of LOOP due to severe weather (SW) piaces the plant in SW Group 2 (Section 3.2.1, Part IC, page 3-7).
: d. The of f-site power system is in the 11/2 Group (NUMARC 87-00, Section 3.2.1, Part 2C, page 3-10).
: 2. The emergency AC (EAC) power configuration group is C based on (NUMARC 87-00 Section 3.2.2, Part 2C, page 3-13):
: a. There are two EAC power supplies not credited as alternate AC power sources (NUMARC 87-00 Section 3.2.2, Part 2A, page 3-15).
: b. One EAC power supply is necessary to operate safe shutdown equipment following a loss of off-site powar (Section 3.2.2, Part 28, page 3-15).
: 3. The target emergency diesel generator (EDG) reliability is 0.95.
A target EDG reliability of 0.95 was selected based on having a nuclear unit average EDG reliability for the last 100 demands (1) Plant-specific data provides an annual frequency of storms with wind velocities greater than or equal to 125 mph of 1 x 10 4; which using                    ,
Table 3-1 in NUMARC 87-00 yields an ESW Group 1.                                        l

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Dr. T. E. Murley                                                                       i B13179/ Attack + 3 No. 1/Page 2                                                      j April 17, 1989 greater than 0.95(2) which is consistent with NUMARC 87-00 Sec-tion 3.2.4.
B.       SB0 Procedure Description
: 1. The following procedures for AC power restoration have been reviewed and modified as necessary to meet the guidelines in NUMARC 87-00, Section 4.2.2:
: a. CONVEX Operating Instruction No. 0006, " Restoration."
: b. CONVEX     Operating    Instruction    No. 6707,    " Connecticut Yankee 12R."
: c. CONVEX Operating Instruction No. 6708, "Haddam Neck 14B."
: d. TD 250, " Load Shedding and Interruptible Loads."
: e. TD 503, " Transmission Line Patrols--Line Faults."                i
: f. TD 506," Transmission Line Emergency Patrols--Regional / Area Assistance."                                                      l
: g. E0P ECA-0.0, " Station Blackout."
: h. E0P ECA-0.1, " Station Blackout Without SI Required."
: i. E0P ECA-0.2, " Station Blackout With SI Required."
: 2. The following plant procedure for severe weather has been reviewed and modified as necessary to meet the guidelines in NUMARC 87-00, Section 4.2.1:
A0P 3.2-5, " Natural Disasters."
: 3. The following plant procedures for SB0 response have been reviewed and procedure changes necessary to meet NUMARC 87-00, Section 4.2.1, will be implemented within one year after NRC notification is provided in accordance with 10CFR50.63(c)(3):
: a. E0P ECA-0.0, " Station Blackout."
: b. E0P ECA-0.1, " Station Blackout Without SI Required."
: c. E0P ECA-0.2, " Station Blackout With SI Required."
(2) The actual average reliability for both EDGs combined for the last 100 demands is 0.99.

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                    . Dr. T. E. Murley B13179/ Attachment No. 1/Page 3 April 17, 1989
: 4. Procedure changes associated with any modifications required after assessing coping capability per NUMARC 87-00, Section.7 have been reviewed.
No modifications are necessary to cope with an SB0 for 4 hours;                                    I however 'a procedure change is necessary as described below for condensate inventory. This procedure change will be implemented within one year after NRC notification is provided in accordance with10CFR50.63(c)(3).
C.                    E 0 Cooina Ascessment The ability of. the Haddam Neck Plant to cope with an SB0 for 4 hours in                                 !
accordance with NUMARC 87-00, Section 3.2.5, and as determined in                                        l Part "A" above, was assessed using NUMARC 87-00, Section 7,                   with the following results:
: 1. Condensate Inventory for Decay Heat Removal (Section 7.2.1).
It has been determined as required by Section 7.2.1 of NUMARC 87-00 that 58,400 gallons of water are required for decay heat removal and cooldown over 4 hours. The minimum permissible condensate level from the demineralized water storage tank per Technical Specifica-                                 h tions provides 50,000 gallons of water. The primary water storage tank, with a minimum capacity per Technical Specifications of 80,000 gallons, and the Connecticut River, via the fire header, are alternate water sources of auxiliary feedwater (AFW).
ECA 0.0 will be modified within one year after NRC- notification is                                '
provided in accordance with 10CFR50.63(c)(3) to include criteria for switching from one source to the next and valve lineups necessary for use of each water source and, in addition, any necessary sup-porting equipment will be procured.
: 2. Class IE Battery Capacity (Section 7.2.2)
A battery capacity calculation has been performed pursuant to NUMARC 87-00, Section 7.2.2,   and verifies that the Class lE bat-teries have suffig Haddam Neck Plant'jgnt capacity to meet SB0 loads for 4 hours at the
: 3. Compressed Air (Section 7.2.3)
Air-operated valves relied upon to cope with an SB0 for 4 hours can either be operated manually or have sufficient backup sources (3) This battery calculation takes into consideration the equipment being                                                      !
added as part of the new switchgear building which will be in service after start-up from the 1989 refueling outage.

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Dr. T. E. Murley B13179/ Attachment No. 1/Page 4
          . April 17, 1989 independent of the preferred and blacked-out unit's Class IE AC power supply. Valves requiring manual operation or that need backup             j sources for operation are identified in plant procedures.                       i
: 4. Effects of Loss of Ventilation (Section 7.2.4)
: a. The assumption in NUMARC 87-00, Section 2.7.1, that the control room at the Haddam Neck Plant will not exceed 120*F during an SB0 has been assessed and determined to be correct. Therefore, the control room is not a dominant area of concern.
: b. The calculated steady state ambient air temperature for the steam-driven AFW pump room (the dominant area of concern) during an SB0-induced loss of ventilation is 162*F.
Reasonable assurance of the operability of SB0 response equip-ment in the above dominant area has been demonstrated using Appendix F to NUMARC 87-00, the NUMARC Topical Report, and plant-specific data. No modifications or associated procedures are required to provide reasonable assurance for equipment operability.
: 5. Containment Isolation (Section 7.2.5)
The plant list of containment isolation valves has been reviewed to verify that valves which must be operated (cycled) under station blackout conditions can be positioned (with indication) independent of the preferred and blacked-out unit's Class lE AC power supplies.
No plant modifications and/or associated procedure changes were determined to be required to ensure that appropriate containment integrity can be provided under SB0 conditions.
: 6. Reactor Coolant Inventory (Section 2.5)
The ability to maintain adequate reactor coolant system (RCS) inventory to ensure that the core is cooled has been assessed for 4 hours. A plant-specific analysis was used for this assessment.
The expected rates of reactor coolant inventory loss under SB0 conditions do not result in core uncovery.      Therefore, RCS makeup systems under SB0 conditions are not required to maintain core cooling under natural circulation (including reflux boiling).
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