ML20012E999

Suppls Info to 890417 Submittal on Station Blackout Rule (10CFR50.63).Instrumentation Required to Cope W/Station Blackout Is safety-related & Evaluated for Operability at Elevated Temps

ML20012E999
Person / Time
Site:	Seabrook
Issue date:	03/30/1990
From:	Feigenbaum T PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NYN-90083, NUDOCS 9004090276
Download: ML20012E999 (7)

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Senior Vice Presided anci Cliief Operating CWiicer 4

NYN- 90083 March 30, 1990  !

United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Document Control Desk i

References:

(a) Facility Operating License No. NPF-86, Docket No. 50-443 (b) Station Blackout Rule (10CFR50.63) j (c) New Hampshire Yankee Letter NYN-89038 dated April 17, ^

1989, 'Information Submittal Required by 10CFR50.63",

G. S. Thomas to USNRC (d) Nuclear Management and Resources Council (NUMARC) Letter N90-005 dated January 4, 1990, ' Station Blackout (SBO)

Implementation: Request for Supplemental SB0 Submittal to i NRC', B. Lee Jr. to NUMARC Board of Directors (e) USNRC Letter dated January 3, 1990, A. C. Thadani to A. t Marion (NUMARC) {

Subject:

Supplemental Information Submittal on Station Blackout Rule {

Gentlemen:

i on April 17, 1989, New Hampshire Yankee (NHY) submitted Reference (c), which provided the evaluation of Seabrook Station Unit 1 against the requirements of the Station Blackout (SBO) Rule, 10CFR50.63. This

• evaluation was based on the use of NUMARC 87-00, " Guidelines and Technical '

Bases for NUMARC Initiatives Addressing Station Blackout at Light Water-Reactors", except where USNRC Regulatory Guide 1.155 takes precedence.

The NRC has performed audits of SB0 submittals of other licensees.

As a result of these audits, the NRC expressed a concern regarding the t departures from the reconunended NUMARC 87-00 methodologies, and the lack of identification of such departures in the April 17, 1989 submittals. NRC expressed their concern in Reference (e), and as a result NUMARC, in Reference (d), requested licensees to supplement their original letter identifying such departures.

NUMARC 87-00 allows plant-specific analysis to be used as ,

supplemental to or in lieu of the recommended methodologies outlined therein. Supporting information is equired to be available where such alternative methodologies are used.

9004090276 900330 PDR ADOCK 05000443 P PDC New Hampshire Yankee Division of Public Service Company of New Hampshire L i P.O. Box 300

• Seabrook, NH 03874

• Telephoee (603) 474-9521

e United States Nuclear Regulatory Commission March 30, 1990

. Attention: Document Control Desk Page two o

In supporting documentation for the original NHY submittal (Reference f (c)), plant specific data or alternative methodologies have been utilized in the areas of estimating the frequency of loss of offsite power due to  ;

severe and extremely severe weather, and assessment of the effects of loss of ventilation. This letter provides supplemental information pertaining to

-these areas. In addition, this letter provides supplemental information or clarification in certain areas of concern identified by the NRC during the j site audits at other plants. .

It should be noted that the results of the evaluation of Seabrook Station Unit I against the requirements of the SB0 rule provided in Reference (c) remain valid and unchanged.

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I. NUMARC 87-00 General Criteria and Base Line Assumptions Section 2.4.1 of NUMARC 67-00 refers to safe shutdown (i.e., hot  ;

standby or hot shutdown as appropriate), following the loss of all l offsite power. NHY provides the following clarification:

• In the event of an SBO, Seabrook Station will proceed with plant I cooldown. The cooldown will continue until the secondary side '

pressure is reduced to approximately 250 psig. The plant will remain in that condition for the duration of the event. .

j Section 2.9.1 of NUMARC 87-00 refers to ' erratic perfor-mance or failure of shutdown instrumentation' during an SB0 event and l the use of backup instrumentation. NHY provides the following L clarifications  !

Instrumentation required to cope with SB0 is safety-related and has been evaluated for operability at elevated temperatures. No l instrumentation will be subjected to temperatures beyond its l L qualification before or during the time that it is required to be  !

I' operable, l y II. Proposed Station Blackout Duration l

l In Reference (c) NHY determined that the estimated frequency of

! lono of offsite power due to severe weather and extremely severe weather, places Seabrook Station in SW Group 3 and ESW Group 3 .

! respectively. NHY clarifies that site-specific weather data and l

supporting calculations were utilized in making these determina-tions.

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i I United States Nuclear Regulatory Commission March 30, 1990 i Attention: Document Control Desk Page three i

Decause Seabrook Station is a coastal plant, the AC Power Design  !

Characteristic Group P2 was determined using Table 3-5b of NUMARC 87-00. 7 r  :

l' The selected target Emergency Diesel Generator (EDG) reliability of '

O.975 will be maintained by implementation of a Diesel Generator  ;

y Reliability Program meeting the Guidelines of USNRC Regulatory Guide 1.155.

111. Procedure Descriution 1

As indicated in Reference (c). Station procedures have been reviewed -

and modified to meet the guidelines of NUMARC 87-00, Section 4. .

Appropriate Station personnel have been trained on the modified  !

procedures.

In Reference (c), NHY indicated that Sb0 responses per NUMARC 87-00, i

Section 4.2.1(1) through 4.2.1(11), are covered in Emergency Contingency Actions (ECA) procedure ECA 0.0 (Loss of all AC Power). '

NHY provides the following clarifications. NUMARC 87-00, Section 4.2.1(12) was not addressed in procedure ECA 0.0 since no Non-1E ,

equipment is required to cope with SBO. Similarly Section 4.2.1(13) of NUMARC 87-00 was not addressed in ECA 0.0 because loss of heat tracing does not affect any equipment required to cope with SBO. The NHY evaluation assumed that underground piping and piping located in permanent structures would be inherently protected from ,

freezing during the relatively short duration of SBO.

Seabrook Station AC power restoration procedures do not address NUMARC 87-00, Section 4.2.2(4) on portable generators, since no .

portable generators are utilired to cope with SBO.

Finally, NHY clarifies that shutdown in anticipation of hurricane arrival at the site is not mandatory, but is at the discretion of plant management [(Section 4.2.3(4)(a)).

IV. Coninn Assensment (A) Battery Capacity The following supplemental information is provided regarding the assessment of the Class 1E battery capacity (NUMARC 87-00 Section 7.2.2):

A battery capacity calculation was performed to verify that the Station batteries have sufficient capacity to cope with SB0 for four hours, assuming that loads not needed to cope are removed. This calculation followed the

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u Unit.ed States Nuclear Regulatory Commission March 30. 1990 Attention: Document Control Desk Page four ,

l' guidelines provided in IEEE 485 for sizing batteries, including load for starting the EDG and closing the t

circuit breakers needed to reconnect the ED3. The l calculation also used the minimum temperato:e anticipated i during the SBO.

(B) Loss of Ventilstion NHY provides the following supplemental information regarding i the assessment of the effects of loss of ventilation. Plant-specific analyses and alternative methodologies are identified where applicable:

Areas containing SB0 equipment were evaluated for the effects of loss of ventilation. These areas include ,

Emergency Feedwater Pumphouse Vital Switchgear Room (A&B) and Battery Rooms

• Containment Structure '

• Main Steam /Feedwater Pipe Chase (East and West)

• Main Steam /Feedwater Pipe Chase Stairwell (West) '

Main Steam /Feedwater Pipe chase Electrical Room (East)

• Mechanical Penetration Area

• Electrical Tunnels

• Main Control Room The methodology of NUMARC 87-00, Section 7.2.4 has limited application for Seabrook Station due to external thermal influences and the complexity of area geometry. In numerous cases, plant-specific analyses are utilized or the methodology .

of Sec. tion 7.2.4 has been modified to suit the specific -

application. The following is a summary of these applications:

1. The Emergency Feedwater Pumphouse was evaluated using the methodology of Section 7.2.4. A plant-specific analysis is also referenced which provides further substantiation of the calculated, steady-state temperature.

2. The Vital Switchgear Room vias evaluated using the L

methodology of Section 7.2.4. A plant-specific analysis is also referenced which provid6s further substantiation of the calculated temperature. The plant-specific scalysis also provides the basis for determining the temperature in the Battery Rooms.

3. The Containment Structure was evaluated using plant-specific analyses other than the LOCA/HELB analyses.

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United States Nuclear Regulatory Commission March 30, 1990 t Attention: Document Control Desk Page five

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These analyses are better suited to deal with the high temperatures and the complex geometry of Containment. The '

analyses assumed that the ventilation fans were not operating.

4. The Main Steam /Feedwater Pipe Chases were evaluated using plant-specific analyses modified to suit the SB0 .

condition. The Saction 7.2.4 methodology is not appropriate for these areas, i

5. The Main Steam /Feedwater Pipe Chase Stairwell (West) was evaluated using the methodology in Section 7.2.4. A .

pinnt-specific analysis, modified for SBO, provides further substantiation of the calculated, steady.. state  ;

temperature.

6. The Main Steam /Feedwater P3pe Chase Electrical Room (East) was evaluated using the methodology of Section 7.2.4, modified to suit this area. A modified analysis was necessary to account for external thermal influences.

7. The. Mechanical Penetration Area was evaluated using the methodology of Section 7.2.4, modified to suit this area.-

/. modified analysis was necessary to account for external thermal influences.  ;

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8. The Electrical Tunnels were evaluated using a plant-specific analysis. The results of this analysis are more -

conservative than the results using the methodology of [

Section 7.2.4.

9. The Main Control Room was evaluated using a plant-specific analysis. This analysis was performed prior to the isbuance of the NUMARC 87-00 guidelines. However, this anklysis specifically addresses the SB0 condition. ,

Continuous drop ceiling and air volumes above have been properly accounted for in the analysis.

(C) Containment Isolation NHY provides the following clarifications regarding Contain-ment Isolation valve listing:

The Containment Isolation valves listed in Seabrook Station FSAR Table 6.2-83 were reviewed per the NUMARC 87-00 criteria for valve identification (Section 7.2.5). Exclusion Item (2)

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of Section 7.2.5 was modified to allow exclusion of valves on loss of AC power only. Exclusion of valves that close on loss of air alone is considered inappropriate since some valves have DC-powered control solenoids, and loss of air alone will not cause such valves to shut. This position is in compliance with USNRC Regulatory Guide 1.155, Parage;aph 3.2.7.

In addition to the five listed exclusion criteria, valves other than those isolated by 6 Containment Isolation Phase 'A", or Phase 'B', or a Containment Purge and Exhaust Isolation sigesl as defined in Technical Specification 4.6.3.2 are not con-sidered valves that provide Containment integrity. Therefore, these valves are not identified by procedures as Containment Isolation valves that may be closed during SBO.

Containment sump isolation valves CBS-V-8 and CBS V-14 are listed in FSAR Table 6.2-83. Specific consideration was given to determining the proper designation for these valves by reasoning as follows. These valves would be in the closed position at all times except during surveillance testing, or in the event of an accident, such as LOCA or steam line/feedwater line break inside Containment. Hence, under accident conditions, when Containment Isolation would most probably be activated, these valves would be in the open position. Also, in accordance with the American National Standard for Containment Isolation, these valves are technically not Containment Isolation valves (Reference ANS 56.2/ ANSI N271-lP76 Page 45, Figure B-8, Note 56-1. Page 36).

furthermore, the PSAR list of Containment Isolation valves (Table 6.2-83) takes exception to the applicable general design criteria for these valves (second colunn of Table 6.2-83).

This is further demonstrated by the fact that these valves do not fit into any of the categories outlined in FSAR Section I 6.2.4.2(d) which defines the applicable criteria for' j For the-abovo identifying Containment Isolation valves.

reasons, valves CBS V-8 and CBS-V-14 have been excluded from the list of valves that may be closed during SBO.

Check valves are excluded from evaluation for operability at elevated temperatures. This position is consistent with and i is an extension of the exclusion criteria of Section 7.2.5, Item (3).

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United Statis Nuclear Regulatory Conunission March 30, 1990 Attentions -Document Control Desk Page seven I

i The information contained in Reference (c) for areas not addressed above was considered adequate and not requiring clarifications or supplemental information, j The evaluation submitted by Reference (c) as well as the j clarifications and supplementary information, provided in this letter are

! supported by technically sufficient documentation available on site.

If you have any questions regarding the above, please contact Mr. John B. Hart, Licensing Manager, at (603) 474-9521, extension 2762. ,

Very truly yours .

y n_ --

Ted C. Feigenbaum t

l cci Mr. William T. Russell Regional Administrator United States Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Mr. Victor Nerses. Project Manager Project Directorato I-3 United States Nuclear Regulatory Commission Division of Reactor Projects  ;

Washington, DC 20555 Mr. Noel Dudley NRC Senior Resident Inspector P.O. Box 1149

'. Seabrook, NH 03874 Mr. Alex Marion Manager, Technical Division Nuclear Management and Resources Council 1776 Eye Street, Northwest Suite No. 300 Washington, DC 20006-2496 l

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