ML20203C939

Forwards Addl Info Per 860407 Request,Including Equipment Operability Time,Emergency Feedwater Sys Mod,Previous Hot Operational Experience of Shift Superintendent & Revised Excerpt from FSAR Section 6.2.1.2

ML20203C939
Person / Time
Site:	Seabrook
Issue date:	04/16/1986
From:	Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:	Noonan V Office of Nuclear Reactor Regulation
References
RTR-REGGD-01.089, RTR-REGGD-1.089 SBN-1006, NUDOCS 8604210268
Download: ML20203C939 (10)

Text

SEABROOK STATION

• Engin:ering office April 16, 1986 g gggg SBN- 1006 T.F. B7.1.2 New Hampshire Yankee Division United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. Vincent S. Noonan, Project Director PWR Project 31 rectorate No. 5

References:

(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-44 3 and 50-444 Subj ect : NRC Requests for Additional Information

Dear Sir:

In discussions with our Bet tenda Licensing Of fice during the week of April 7, 1986, the Staf f requested additional inf ormation/clarifica-tions concerning a few items. In response to these questions, enclosed please find the following:

o Further information concerning EQ operability time to that previously submitted by PSNH letter dated April 3,1986

( SBN-988) - At tachment I o Further information relative to the EFW system modification that was previously submitted by PSNH letter dated April 1, 1986 (SBN-984) - Attachment 2 o Additional information concerning previous hot operational experience of Shif t Superintendent - Attachment 3 o Revised excerpt of FS AR Section 6.2.1.2 (p 6.2-22) -

Attachment 4 We trust that the enclosed provides the additional information/clari-fications requested by the Staf f and request that the acceptability of the enclosed, where applicable, be reflected in the upcoming supplement to Seabrook's SER.

Very truly yours,

/

B604210260 860416 , ./ w PDR ADOCK 05000443 John DeVincentis, Director A PDR Engineering and 1,1 censing Enclosures cc: Atomic Safety and 1,1 censing Board Service List i

Smbrook Station Construction F!o!d Offico . P O Box 700 Soobrook, NH 03874

_ _ . _ _ _ ._ _ _ _ . _ _ _ _ _ _ ___ _ . _ _ _ __ _m . _ __ __ _

s

• j Diane Curren B; quire Colvia A. Canner .

Harmon & Weiss City Manager 2001 S. Street, N.W. City Hall Suite 430 126 Daniel Stredt '

Washington, D.C. 20009 -

Portsmouth, NN 03801 Sherwin B. Turk Esq. Stephen E. Merrill, Esquire Office of the Executive Legal Director Attorney Ceneral U.S. Buclear Regulatory Commission George Dana Bisbee, Esquire Tenth Floor Assistant Attorney Ceneral Washington, DC 20555 Office of the Attorney Ceneral 25 Capitol Street

Robert A. Backus Esquire Concord, EN 03301-6397 116 Lowell Street P.O. Box 516 Nr. J. P. Nedeau l Nancheeter, MN 03105 Seleetaen's office l 10 Central Road Philip Ahrens, Esquire Rye, MN 03870 Assistant Attorney General i Department of The Attorney General Mr. Angle Nachiros Statehouse Station M Chairman of the Board of Selectmen hugusta, IE 04333 Town of Newbury Newbury, MA 01950 Nrs. Sandra Cavutte Chairman, Board of Selectmen Mr. William S. Lord RFD 1 - Box 1154 Board of Selectmen a Resuutngton, MN 03S27 Town Hall - Friend Street l Amesbury, MA 01913

, Carel S. Sneider, Esquire

! Assistant Attorney General Senator Gordon J. Numphrey j Department of the Attorney General 1 Fillsbury Street One Ashburton Place,19th Floor Concord, EN 03301 ,

Boston, MA 02108 (ATTN Norb Soynton)

Senator Gordon J. Numphrey M. Joseph Flynn, Esquire U.S. Senate Office of General Counse! ,

Washington, DC 20510 Federal Baergency Management AgencFt (ATTW: Tom Durack) 500 C Street, SW.

Washington, DC i20472 Richard A. Hampe, Esq.

Nampe and Mc51cholas Paul McEachern, Esquire 35 Pleasant Street Matthew T. Brock, Esquire Concord, BN 03301 Shatnes & Nctachern

! 25 Naplewood Avenue Donald E. Chick P.O. Box 360

Town Manager Portsmouth, NN 03801  ;

Town of Exeter  :

10 Front Street Gary W. Holmes Esq. '

Exeter, WM 03833 Holmes & Ells 1 47 Winnacunnet Road Brentwood Board of Selectmen Hampton, WH 03041 RFD Dalton Road l

Srentwood, BN 03033 Nr. Ed Thomas FEMA Region I ,

Peter J. Mathews, Mayor 442 John W. McCormack PO & Ccurthouse City Hall Boston, MA 02109 j gewburyport, MA 01950 l Stanley W. Knowles, Chainnan Soard of Selectmen I P.O. Box 710 4 North Hampton, NH 03062

1 Administrative Judge Helen Hoyt, Chairperson Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Administrative Judge Sheldon J. Wolfe, Chairman Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dr. Eameth A. Luebke Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dr. Jerry Harbour Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 l

SBN- 1006

• ATTACHMENT 1 POST ACCIDENT OPERABILITY TIME, ADDITIONAL INFORMATION The following is a comparison of the Engineering justification submitted in SBN-Sd8 with Regulatory Guide 1.89 requirements for equipment required to perform its safety function only within the first ten hours of an event.

For the Feedwater Isolation Valves R.G. 1.89 recommends that the equipment remain functional in the accident environment for a period of at least I hour in excess of the time assumed in the analysis. Our analysis conservatively indicated that valve closure will occur in 1 minute and qualified post accident operating time is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> which meets the recommendation. R.G. 1.89 requires a spectrum of breaks be considered. We have considered a spectrum of break sizes based on the design requirements for "superpipe" which is the type of pipe where the valves are located. The valve closure time was based on the smallest break and the environment based on the largest break. R.G. 1.39 re-quires consideration for the need of the equipment later in the event or during recovery operations. We stated that the plant would be safely shut down using the Auxiliary Feedwater System, which does not require operation of the Feedwater Isolation Valves (FIV). R.C. 1.89 requires that failure of the equipment af ter performance of its safety function will not be detrimental to plant safety or mislead the operator. We stated that Control Room indication is provided by Namco 180 Limit Switches which are qualified for 1 year post-accident operation. Any eq ui pment failure causing the FIV to open would not be detrimental because of the F.W. check valves located immediately upstream. R.G. 1.89 requires a determination that the time margin applied to the minimum operability time account for uncertainties associated with the analysis. We con-cluded the four hour post-accident operability qualification has suf ficient margin, and in fact surpassed the time margin required by lEEE 323-1974.

A similar comparison to R.G. 1.89 for the ASCO Temperatures Switches (File 252-38-01) is not practical because the functional requirement is greater than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> and the qualified post-accident qualification time is 30 days. The worst case spectrum of events considered for the PAB lo-cation of the temperature switches last for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> therefore, the 30 day post-accident qualification is more than sufficient. The peak temp-erature is 135'F. As stated previously in SBN-988 the need for this equipment is to automatically start safety grade f ans (Pall-FN-42A&B). In the event of a temperature switch f ailure the f ans have a continuous run feature which will override the temperature switch f rom control roon operation. The fans have Class IE Control Room operational indication which is direct f rom the fan motor cont rol center. A separate high temperature alarm is available to the Control Room in the event of a temperature switch f ailure. We have determined that the margin for min-imum operability time is suf ficient and nects the requirement for IEEE 323-1974.

SBN-1006 o

ATTACHMENT 2 EFW SYSTEM MODIFICATIONS, ADDITIONAL INFORMATION As a result of a telecon discussion between D. A. Maidrand. P. H.

Hannes, and NRC NRR personnel (04/09/86), the NRC requested clari-fication to specific concerns regarding the Seabrook Emergency Feed-water sys tem. The following documents the concerns / questions addressed and provides the appropriate response and resolution.

QUESTION NO. 1 Why didn't MSV-127 and MSV-128 readily open up)n receipt of an EFW initiation signal? Clarif y how MSV-127 and MSV-128 opened during Hot Functional Testing.

RESPONSE

The f unctional reason why MSV-127 and MSV-128 didn't readily open is due to the application of these type valves a s EFW steam supply iso-lations. MSV-127 and MSV-128 are wedge-type ga te valves (f ail-open).

During an EFW system standby condition, there is approximately a 1100 psi di f ferential across these valves. This high dif ferential pressure across the valve (s) gate contributed to a lagged opening response.

During Hot Functional Testing, it was observed that MSV-127 and MSV-128 didn't simultaneously open upon EFW system initiation. Die isolation valve opened followed by opening of the alternate isolation valve shortly thereaf ter.

OUESTION NO. 2 Why will the new EFW isolation valves (MSV-393, MSV-394) alleviate the ope:11ng problem that existed with MSV-127 and MSV-1287

RESPONSE

MSV-127 at MSV-128 are no longer utilized as EFW steam supply isolation val ve s. The ,e valves have been reassigned a " normal open" position. The new EFW s team isolation valves (MSV-393, MSV-394) are globe valves. In the shut position, a steam header pressure of approximately 1100 put is applied beneath the valve (s) disk. This steam pressure aids the opening of MSV-393 and MSV-394 upon receipt of an EFW initiation signal.

i QUESTION No. 3 k

k When will plant operators take action to close MSV-127 and/or MSV-1287

RESPONSE

l l

a. Steam Generator Tube Rupture Event if necessary.

Sr

b. Valve surveillance testing.

Note: At no time should both MSV-127 and MSV-128 be closed concurrectly. <

l I QUESTION NO. 4 When will the EFW pump (s) recirculation line MOVs be opened?

RESPONSE

Applicable Seabrook Station Procedures are being updated to reflect l opening the EFW recirculation line isolations valves prior to throt ti-ing EFW flow to the SGs. This will ensure the minimum required EFW pump (s) recirculation fl ow. The attached excerpts (pgs. 6&7) are i typical of procedural provisions being invoked to ensure the minimum l EFW pump (s) recirculation flow is retained during EFW system operation.

ATTACTIMENT 2 (continued) SBN-1006

Code

Symptom / Tit e: Ef"C"d"F" N"'

e Revision No./Date g hu i1f, N/A SAFE SitUTUUWN AND '.OULDOWN Fh0M 'IttL hr.hu t c. a r r. US1200.02 SHUTDUWN FACILITIES 00 /

l STEP l l AGTION/LXPECTt.D RESPONbE l l RESPONSE 50T OtlTAINED l _ _..

6 Check SG Pressures:

CP-lU6A CP-lubb SG-A PI-3173 SG-B PI-3174 SG-C PI-3178 SG-D PI-3179

a. Control SG atmospheric steam a. IF,RSS control panci control dump valves to - MAINTAIN is ineffective, take local 1100 PSIG UNTIL C00LDOWN control of atmospheric COMMENCES: steam dump valve (s).

e Jog Switches

- OR -

e Modulating Control CP-106A CP-106B J0G SW JOG SW SG-A HIC-3OU1 SG-B HIC-3OUZ J0G SW J0G SW SG-C HIC-3003 SG-D HIC-3004 CAUTION OPEN EFW pump recirculation valves be 3re throttling EFW Flow.

7 Check SG WR Level:

l CP-108A CP-108B SG-A LR-4310, LI-4310 SG-B_ LR-4320, L1-4320 SG-C LR-4310, LI-4330 SG-D LR-4320, L1-4340 *

a. Control EFW flow control a. JF designated EFW flow control l valves to maintain SG WR valves do NOT control flow, l level - BETWEEN 65% AND use valves on opposite train 90%: RSS panel , r!R cont rol locally.

i

~

l CP-lONA _ C P .l u n n _._

~ ~ E~1'O ._C_ _ E_ '. -

~.... _ _CP._10.N_A _ ._..

SG-A FW-FV-4214A SG-B FW-FV-4224t$ SG-A FW-FV-4214H SG-li FW-FV-4/24A l SG-C FW-FV-4214A SG-D FW-FV-4244n SG-C FW-F V -4 214 H SG-D FW-FV-4244A

, 79l,

ATTACHMENT 2 (continusd) SBN-1006

~~

Symptom /Ti tle Procedure No.

Kevision No./Date:

SAFE SIIUTuv.3 A3u tuvLeva.. r nv., i ., a .iab CUM a0L 0S1200.01 Roon 00 /

l STEP l l ACT10N/ EXPECTED RESPONSE l lRESPONSENOTOBTAIN,E_D,[_

8 Perform Applicable Manual Disabling Functions Based On i Fire Area / Zone - APPENDIX A WHILE CONTINUING WITH THIS PROCEDURE ,

9 Check SG Pressures: ,

a Control available SG ASOVs - a. Control SG ASDVs in jog mode MAINTAIN 1100 PSIG UNTIL - OR -

C00LDOWN COMMENCES Take local control of SG ASDVs.

CAUTION: Open EFW Pump Recirculation Vaive(s) Bef ore Throttling EFW Flow.

10 Check SG WR Level:

i a Maintain SG 1evel: a. Control EFW flew con".rol valves as necessary.

i BETWEEN 5% AND 50% NR f - OR -

BETWEEN 65% AND 90% WR l 11 Check RCS Pressure:

I a Maintain PRZR pressure - a. JF designated heater groups BETWEEN 2100 AND 2250 are inef f ective, THEN use PSIG the following:

• Proportional heater group e Backup heater group A e Backup heater group C e Backup heater group B l

e Backup heater group D

b. IF, pressure continues to decrease THEN verify PORVs closed OR closu block valves.

- MD -

f Verify Aux spray isolated: i l e CS-V185 - CLOSED

- OR -

l e CS-V!42 - CLOSt.D

- OR -

e Cb-V143 - CLOSED 6 of 2)

I l

f

_m___.___._.____

4 ATTACIMENT 3 SUN loo 6 .

Seetrcok Station SHIFT SUPERINTENDENT Previous Experience

( Months)

- EDUCATION - MlUTARY COMMERCIAL - OTHER EXPERIENCE noT OPS

• NUCLEAR Test Asiated EXP. .

PREVIOUS Catega NUCLEAR Exp Ro/SRO AGE NRC UC. Creets Degree EXPERIENCE EXPERIENCE Reactor Fossi NAME 57 6/1 (ewa)

D:vid, MR 39 RO/SRO BS 58EWS BS 69ERS 72 49/o rewa)

Fritz, LH 38 RO 144 36 30/11 (ewa) 52 RO/SRO AS Madel EV 69 55/o (ewa)

Peterson,JL 39 RO 176 38EWS 24 ERS 81 41/o (awa)

Strickland, RG 38 RO BS BS 80EWS 61 13/o (pw3)

Thompson, RB 40 RO Statt 118 28 EOOW 114 28 48/72 (ewa)

~ Walsh, LA 45 RO/SRO 104 48/12 (eway Grillo, JM 39 RO/SRO BS, AS 36ERS BS 72 EOOW 40 o/o Malone, JM 40

SEM- 1006 6

• Abcknwk 4 SB 1 & 2 Amendment 54 FSAR February 1985

3. Pressurizer Compartment The pressurizer coupartment is a reinforced concrete structure extending f rom El. O'-0" to El . 63'-0" which encloses the pressurizer and its associated piping. The pressurizer skirt, which is a cylindrical support extending from the bottom of the pressurizer, anchors the pressurizer to the compartment floor. A ring support at E1. 23'-6 3/4" provides lateral support for the pressurizer. Section and plan drawings of g the pressurizer compartment are shown in Figures 6.2-35 through 6.2-40. Free volumes and vent areas have been calcu-lated assuming the insulation remains intact during the transient. The llVAC ducting and sheet metal panels at eleva-tion 16'-6" are designed to bicv out in the event of a pressure buildup of 0.25 psig in the compartment to prov'de additional vent area. The total free volume of the compartment used in the analysis is 6638 cu. ft. and the total vent area to the containment is 400 sq. ft.

41

4. Pressurizer Skirt Cavity The pressurizer skirt cavity is formed by the bottom of the ,

pressurizer and its supporting skirt. A 14" surge line which connects the reactor coolant system with the pressurizer passes through a 5b ft. diameter opening in the pressurizer compartment floor. Figures 6.2-41 and 6.2-42 show the plan and elevation drawings of the pressurizer skirt cavity. The volume below the pressurizer skirt has a large vent opening to the containment. The total free volume of the skirt cavity is 1860 cu. ft. and the total vent area to the containment is 238 sq. ft. The insulation on the pressurizer and on the surge line is assumed intact in calculating the free volume and vent openings.

c. Design Evaluation

1. Mass and Energy Release Data The mass and energy release data for all the breaks conside-ed for the subcompartment analyses has been generated by Westfog-house. Discussions of the blowdown model are provided in Reference

2. Computer Code for Subcompartment Pressuritation Analysis The subcompartment pressure transients were calculated using COMPRESS - a digital computer program. A detailed description l of the analytical mett.od can be found in Appendix 15C and 54 Reference 2. Some important aspects of the method are outlined belows

(

6.2-22