ML20094N672

Forwards FSAR Commitment Status Through June & Jul 1984 & Responses to NRC Request for Addl Info Re FSAR Amend 7

ML20094N672
Person / Time
Site:	Hope Creek
Issue date:	08/13/1984
From:	Mittl R Public Service Enterprise Group
To:	Schwencer A Office of Nuclear Reactor Regulation
References
NUDOCS 8408160336
Download: ML20094N672 (51)

Text

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PutracS2rwce O PS G Company Electnc and Gas 80 Park Plaza, Newark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation August l3, 1984 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda,. Maryland 20814 Attention: Mr. Albert Schwencer, Chief Licensing Branch 2 Division of Licensing Gentlemen:

HOPE CREEK GENERATING STATION DOCKET NO. 50-354 FSAR COMMITMENT STATUS THROUGH JULY 1984 Public Service Electric and Gas Company presently plans to issue Amendment No. 7 to the Hope Creek Generating Station Final Safety Analysis Report in August 1984. This letter is provided, along with three (3) signed originals of the required affidavit, to document the status of Hope Creek Generating Station responses to NRC requests for additional information which were forecasted to be responded to by June and July 1984.

Attachment I is a tabulation of the Hope Creek Generating Station Final Safety Analysis Report commitments for June 1984, and the corresponding resolution for each commitment.

Attachments II through IX provide the responses to the questions forecasted to be responded to in June 1984, which will be included in Amendment No. 7.

Attachment X is a tabulation of the Hope Creek Generating Station Final Safety Analysis Report commitments for July 1984, and the corresponding resolution for each commitment.

Attachments XI and XII provide responses to questions forecasted to be responded to in July 1984, which will be included in Amendment No. 8.

8408160336 840813 i PDR ADOCK 05000354 A A PDR ,1)gD \

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The Energy People ,

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DirectorJof Nuclear

. Reactor Regulation 2 -8/ 13/84 Should you'have anyfquestions in this regard,.please contact cus.

Very truly.yours, YJ Attachment 11- - Hope Creek Generating Station -'FSAR Commitment Status-through' June 1984 Attachment II. - Response to Question 210.12 Attachment III -- Response to Question ~220.15

. Attachment IV. - Response 7t o-Question 281.2

= Attachment V: -Response to Question 410.38 Attachment VI .- Response to Question 410.39 Attachment VII - Response to Question 410.42 Attachment VIII.--Response to Question 421.13 Attachment IX. - Response to Question 630.7e&f Attachment X - Hope Creek Generating Station - FSAR Commitment Status through July 1984 Attachment XI - Response to Question 210.12 Attachment XII - Response to Question 430.19 C D. H. Wagner (w/ attach)

USNRC Licensing Project Manager W. H. Bateman (w/ attach)

USNRC Senior Resident Inspector MP84'123/06 1/2B

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-354 PUBLIC~ SERVICE ELECTRIC AND GAS COMPANY

- Public Service Electric and Gas Company hereby submits the enclosed Hope Creek Generating Station Final Safety Analysis Report.(FSAR) question responses.

-The matters. set forth in this submittal are true to the best of my knowledge, information, and belief.

Respectfully submitted, Public Service Electric and Gas Company By: M /

Thdhias J.fartin Vice Pre % dent -

Engineering and Construction Sworn to and subscribed before me, a Notary Pubjic of New Jersey, this /h day of August 1984.

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. _ DAYl0 K. BURD NOTARYPUBUC OF NEW JERSEY My Comm. Empires 10-23 85 NM 19 li

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Page 1 of 6 ATTACHMENT I HOPE CREEK GENERATING STATION FSAR COMMITMENT STATUS THROUGH JUNE 1984 FSAR COMMITMENT LOCATION COMMITMENT RESOLUTION ,

1. Question / Response Re: TMI-Item I.C.5: This Appendix: commitment concerns assuring Question 100.6 feedback of operating experience to1 operational personnel via procedures.

This information will be provided in August 1984.

Re: TMI Item II.B.3: This commitment concerns assuring compliance of the radioactive gas and liquid sampling sys-tem for shielding and source term requirements. This information will be provided in August 1984 and September 1984.

2. Question / Response This commitment concerns com-Appendix: pliance with conditions in Question 210.12 conditionally approved Code Cases identified in Reg.

Guides 1.84 and 1.85. This information will be submitted in three parts to be provided in July 1984, October 1984, and November 1984. These revised commitment dates are provided in Attachment II and will be included in Amend-ment 7 to the HCGS FSAR.

3. Question / Response This commitment concerns Appendix: detailed procedures for seis-Question 220.10 mic instrumentation inservice surveillance program. This information will be provided in April 1985. This revised commitment date will be included in Amendment 8 to the HCGS FSAR.

M P84.123/07 1-srd

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Page 2 ofL6.

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FSAR' COMMITMENT LOCATION' COMMITMENT RESOLUTION

4. - Question / Response- ,This commitment' concerns

. Appendix ' -mathematical models.used in QuestionE220;15- the ' design _ of Spent Fuel-

-Racks. ' This '.inf ormation will-be provided in~ September 1984. This/ revised: commit-ment.date is provided in Attachment III-and will be-included.in Amendment 17 to the-HCGS FSAR.

5. : Question / Response:

~

This-commitment concerns the-Appendix: Spent-Fuel Racks and their.-

QuestionJ220.16 conformance with subsection NF of the ASME Code. This information is provided in Amendment 26ito the HCGS FSAR.

6. Question / Response- This commitment concerns

_ Appendix: revising FSAR Section l.8 to Question 260.15 reflect.conformance with listed Reg. Guides which are applicable during operations phase. _This.information will be provided in August 1984.

7. Question / Response This commitment concerns pro-Appendix: viding FSAR Figures 9.1-3 and Question-281.2 9.1-4 which. illustrate Spent Fuel Rack design and arrange-ment. This information is provided in Attachment IV and will be included in Amendment 7 to the HCGS FSAR.

, 8. . Question / Response This commitment concerns Appendix: limits for' dissolved and sus-Question 281.9 pended solids in purified condensate. This information will ~ be provided in December 1984. This revised commit-ment date will bo included in Amendment 8 to the HCGS FSAR.

M P841123/07 2-srd

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Page 3 of 6-FSAR COMMITMENT LOCATION COMMITMENT RESOLUTION

9. Question / Response _ This commitment concerns Appendix: .

chemistry sampling for the Question 281.11 ' Spent Fuel Pool Cleanup Sys-tem. This information will be provided in September 1984. This . revised commit-ment date will be included in Amendment 8 to the HCGS FSAR.

10. Question / Response This commitment concerns the Appendix: materials' monitoring program-Question 281.14 for the Spent Fuel-Pool.

This information will be pro-vided in August 1984.

11. Question / Response This commitment concerns Appendix: information on the Post Acci-Question 281.15 dent Sampling System which demonstrates compliance with NUREG-0737, Item II.B.3.This information will'be provided in August 1984 and September 1984.

12. Question / Response This commitment concerns Appendix: information requested in Question 410.26 G.L. 81-34 regarding BWR Scram System Piping. This information will be provided within 60 days of NRC accep-tance of the BWROG position.

This revised commitment date will be included in Amendment 8 to the HCGS FSAR.

13. Question / Response This commitment concerns the Appendix: criticality of the Spent Fuel Question 410.38 Pool. This information will be provided in September 1984. This revised commit-ment date is provided in Attachment V and will be included in Amendment 7 of the HCGS FSAR.

-M P84 123/07 3-srd

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Page 4 of 6 FSAR COMMITMENT LOCATION COMMITMENT RESOLUTION

. l .4 . . Question / Response This commitment concerns Appendix: . Spent Fuel Rack design de--

Question 410.39 ' tails. .This information is provided-in' Attachment-VI and will be included in Amendment 7 of the HCGS FSAR.

15. Question / Response This commitment concerns the Appendix: highest anticipated assembly Question 410.42 average enrichment of U235 g used in the Spent Fuel Rack criticality analysis. This information is provided in Attachment VII and will be included in Amendment 7 of the HCGS FSAR.

16. Question / Response This commitment concerns the Appendix: ability of check valves in Question 410.91 the Equipment and Floor Drain System to maintain a func-tional pressure boundary.

This information will be provided in August 1984.

17. Question /Rosponse This commitment concerns Appendix: seismic qualifications of Question 410.93 check valves in drainage sys-tems. This information will be provided in August 1984.

18. Question / Response This commitment concerns Appendix: testing of I&c . isolation sys-Question 421.13 tems against the' effects of EMI per IEEE 472-1974 and PMC 33.1-1978. This testing has been completed and this information is provided in Attachment VIII and will be included'in Amendment 7 of the HCGS FSAR.

M P84 123/07 4-srd

PageJ5 of 6 FSAR COMMITMENT.

LOCATION COMMITMENT-RESOLUTION v

19. : Question / Response This commitment concerns Appendix:- reactor. mode switch contact Ouestion 421.26- _misoperations. This informa-tion will be provided in August 1984.

120. Ouestion/ Response This commitment concerns trip Appendix: settings for the plant leak Ouestion 440.10 detection system. This information will be provided-in August 1984.

21. Question / Response This commitment concerns the Appendix: implementation of acceptance Question 460.16 criteria for the licensed solid waste burial facility.

This information will be pro-vided in March 1985. This revised commitment date will be included in Amendment 8 to the HCGS FSAR.

22. Question / Response This commitment concerns pro-Appendix: viding the resume for Senior Question 471.14 Radiation Protection Super-visor. This information will be provided in August 1984.

23. Question / Response This commitment concerns seg-Appendix: ments of the plant training Question 630.7e&f program. This information is provided in Attachment IX and will be included in Amendment 7 of the HCGS FSAR.

24. Supplementary Request This commitment concerns pro-for Additional viding a master listing of Information (5) seismic and dynamic qualifi-cation summary and status of safety related equipment.

This information has been submitted (Refer to: R. L.

Mittl (PSE&G) to A. Schwencer (NRC), dated July 5, 1984).

M P84 123/07 5-srd c

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Page 6 of 6 FSAR' COMMITMENT-

' LOCATION COMMITMENT RESOLUTION

.25., FSAR Tab'le 13.1 This commitment. concerns pro-viding resumes for Main-tenance Manager and Senior Nuclear Maintenance Super-visor. This information will be provided in August 1984.

M.P84 123/07 6-srd

. Attachment II HCGS FSAR 1/84 1

QUESTION 210.12 (SECTION 5.2)

Table 5.2-2 identifies certain ASME Code Cases that have been used in the construction of components for the Hope Creek Generating Station. A number of these Code Cases are identified in Regulatory Guides 1.84 and 1.85 as conditionally acceptable.

That is, Regulatory Position C.1 of each guide identifies additional conditions that should be imposed in addition to those conditions specified in each Code Case. The Code Cases that are identified in Regulatory Guide 1.84 as conditionally acceptable are: 1636/1636-1, 1711, 1734, 1818, N-192, and N-275. The Code Cases that are identified in Regulatory Guide 1.85 as conditionally acceptable are: 1644/1644-3, 4, 6, N-71-9, and N-249. Code Case N-253-1 which is not in the Regulatory Guides is also conditionally acceptable.

Demonstrate that you are in compliance with the additional conditions applicable to each of the above conditionally approved Code Cases that are identified in Regulatory Guides 1.84 and 1.85.

RESPONSE

Code Case 1636-1 (N-70) was invoked in the design of the safety auxiliaries cooling system (SACS) pumps. Regulatory Guide 1.84 states that the design guidance in this code case is acceptable subject to the restriction that the stress limit designations of

" Upset," " Emergency," and " Faulted" should be established and justified in the design specification.

HCGS complies with this additional regulatory requirement. The loading combinations are specified in the design specification.

Code Case 1711 (N-100) was invoked in the design of a number of safety-related safety-relief valves. R'egulatory Guide 1.84 states the design guidance in this code case is acceptable subject to the requirement that the FSAR demonstrate how the pressure relief function is assured if the stress limits utilized for the upset operating condition are in excess of those specified in the code case.

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• m a!!ib i ree7mmem m411 wo 7 7..ig:3 ;- yune 4971, Code Cases 1734 and 1818 were invoked in the fabrication of certain pipe supports. Regulatory Guide 1.84 states that the design guidance in these code cases is acceptable subject to the additional welding restrictions found in the regulatory guide.

HCGS complies with these additional regulatory requirements. The 1 applicable design specifications permit the use of Code Case 1734 '

210.12-1 Amendment 4 l g,---

Attachment II (cont'df HCGS FSAR 1/84 and/or_1818 subject to the limitations recommended by Regulatory Guide 1.84.

Code Case N-192 was invoked in the fabrication of certain

. flexible metal instrument hose assemblies and on certain standby diesel generator skid-to-facility connectors. Regulatory Guide 1.84 states that this code case is acceptable subject to the requirement that the applicant should provide design data to demonstrate compliance with Paragraph NC/ND-3649.

,((ol)~ Information to comoly with this additional regulatory requirement will is com. be provided dead -+=inhrre b pr1984 pr-A o uMr seprv4( CottX. %I.sihf.<m-f:%

CodeCaseN-275wasinvokedinthh'fabricationofcertainsafety-related pipe. Regulatory Guide 1.84 states that the design guidance in this code case is acceptable subject to the additional welding restrictions in the regulatory guide.

HCGS complies with these additional regulatory requirements. The HCGS piping design specification permits the use of Code Case N-275 subject to the limitations recommended by Regulatory Guide 1.84.

Code Cace 1644 and.its various revisions has been invoked in numerous applications. Regulatory Guide 1.85 states that this code case is acceptable subject to the limitations on maximum ultimate tensile strength and, in the case of Code Case 1644-9 ,

(N-71-9), the additional requirements for electrode dispersal.

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Use of Code Case 1644-9 (N-71-9) is subject to the additional precautions cited in Regulatory Guide 1.85.

Use of Code Case N-249 is permitted for the containment hydrogen recombiner technical specification. To date, this code case has not been invoked. .

Code Case N-253-1 provides rules for the construction of ASME components which experience elevated temperatures. This code case recombiners.

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• Attachment III l HCGS FSAR '10/83 OUESTION 220.15 (SECTION 3.8.4)

Provide sketches of the mathematical models used in the design of spent fuel racks. Describe in detail, the methods of analysis by -

which seismic and other loads are applied to the racks and the pool.

. RESPONSE i The requested information will be avail'able by 1984, and will be added to Section 3.8.4 and/or 9.1.2 as appropriate.

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OUESTION 281.2 (SECTION 9.1.2)

Figures 9.1-3 and 9.1-4 are shown as to be supplied later. .

Supply these figures or a date by which they will be provided.

RESPONSE . . .

seeHon 1.1.2 54 heen tevued to fMUc.l2 .

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SPENT FUEL MACK ARRANGEMENT IN FUEL POOL FIGURE 9.14 0

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Attachment V HCGS FSAR .10/83 i

('" bOESTION 410.38 (SECTION 9.1.2)

Insufficient information is provided for review of the criticality of the spent fuel pool. The design bases are i acceptable with respect to criticality. The information required for the review is promised for later. Such information should include the followings

a. Sufficient structural detail to permit an independent calculation of the criticality of the racks.

b. A description of the calculational methods used along with the results of the verification of the methods. This may be by reference to documents previously submitted by the >

organizations doing the analysis.

c. A tabulation of the nominal value of k effective of the

racks along with the various uncertainties and biases considered in the analysis. -

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d. A tabulation of the reactivity effect of each of the ,

. abnorsal (accident) situations considered.

( RESPONSE ,

Sufficient information for review of the criticality of the spent fuel pool, including that listed above will be available by '

4hutt'1984, and will be added to Section 9.1.2.

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Attachment VI BCGS FSAR 10/83

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OUESTION 410.39 (SECTION 9.1.2)

Without the spent fuel storage rack design details, the results of an analysis of impact onto the racks and the bundle-to-bundle fuel spacing, the staff cannot make any favorable conclusions about the design. Provide this information in the FSAR.

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.' Attachment VI (cont ' d)

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( HCGS FSAR 1/84 CHAPTER 9 l l

FIGURES Fioure No. M i 2 9.1-1 New Fuel Rack Arrangement 9.1-2 General Arrangement of Spent Fuel Storage Pool 9.1-3 A Typical Spent Fuel Rack O. d C ;;..tri: ?. 1 ?::it!e-i- F .

a

.: Od"  ?;:1 Ster:d 17. Cer.t ;! 5:e 5::i; " #

^ Air.orcr! 'rr! Siv6 ;; "- '---"'

. Of 9.1-4 Spent Fuel Rack Arrangement in Fuel Pool 9.1-5 Fuel Pool Cooling and Torus Water Cleanup, P&ID 9.1-6 Fuel Pool Filter Domineralizer, P&ID 9.1-7 Fuel Preparation Machine Shown Installed in Fuel

Pool 9.1-8 New Fuel' Inspection Stand 9.1-9 Channel Bolt Wrench 9.1-10 Channel Handling Tool f,

9.1-11 Fuel Pool Sipper i, 9.1-12 Channel Gauging Fixture

9.1-13 Fuel Grapple 9.1-14 General Purpose Grapple

9.1-15 Fuel Inspection Fixture i

i 9.1-16 Refueling Outage Flow Diagram ( 9.1-17 Plan View of Refueling Floor During Refueling j

~

9.1-18 Simplified Section of New Fuel Handling Facilities (Section X-X, Figure 9.1-17) i

+

~ .

i i

'gg Attachment VI (cont'd) 4 i 1. Normal storage conditions e.tist when the fuel storage racks'are located'in the pool and are covered with about 25 feet of water for radiation shielding, and with the maximum number of fuel assemblies or bundles _.in their design storage position.

=: (

2. An abnormal storage condition may result from accidental dropping ~of an' empty fuel rack, or from desags caused by the horizontal movement of fuel haniling.*quipment without first disengaging the fugl-from s s, the hoisting equipment. i

b. It is assvoed that the storage array is infinite in all directions.' Since no credit is taken for leakage, the values reported as effective neutron multiplication factors are'An reality infinite neutron multiplication  !

factors. The biases between the calculated results and esperimental results and the uncertainty involved in the calculations, as well as other uncertainties, are taken into account as part of the calculational procedure to ensure that the specified K,gg limits are ,

1 - met.

lr .' -

l c.- The racks'are designed to pec'tect the fuel assemblies  !

from physical. damage caused ~by impact from fuel assemblies. The rack design would prevent the release of radioactive materials in excess of 10 CFR 20 and  !

i 10 CFR 100 allowances under normal and abnormal storage conditions. l

d. The racks are constructed in accordance with the QA requirements of 10 CFR 50, Appendia B.

e. The spent fuel storage racks ar,t constructed in  !

accordance with Seismic Category I requirements. The  !

applicable code for the design of racks is ASNE l l Section II,1, Subsection NF.

I

f. Spent fuel storage space is provided in the fuel storage' pool to accommodate :t 1 2:t ".;" core loads of I,3 1' fuel assemblies. .

I

,; 1

i. t ' , 9.1-7

}

- _ . , . _ . _ . _ , _ _ - _ , , . _ . . - - _ _ _ . . . . _ - . _ - . _ , _ . _ _ . . _ . _ _ , _ . . m

• * *" I I "

dk2/13 RCGS FSAR I g. Spent fuel storage racks are designed and arranged so ttat the fuel assemblies can be handled efficiently -

during refueling operations.

b. The spent fuel storage facility and all piping connections are designed to prevent a loss of cooling water from the spent fuel pool that could uncover the stored fuel.

1. The spent fuel storage facility is designed to prevent criticality of stored fuel under adverse environmental '

and postulated fuel handling accident conditions.

j. Shielding for the stored spent fuel assemblies is designed to protect plant personnel from esposure to direct radiation greater than that permitted for continuous occupational exposure during normal operations. ,

1

k. The spent fuel storage facility is designed to remain

functional during and following a safe shutdown earthquake (SSE).  ;

i Failures of systems or structures not designed to i 1.

Seismic Category I standards and located in the l vicinity of the spent fuel storage facility do not increase the K,gg established by design.

m. The spent fuel pool is designed to withstand thermal stresses resulting from the pool water boiling.

n. The rack design prevents accidental insertion of fuel

' assemblies between adjacent racks.

I.

o. The spent fuel storage facility is designed so that l l

failure of structures, systems, or components that are

' not Seismic Category I will not result in a loss of function of the facility.

P. The speaf N.I :f 1411F entwAMeat $.=V isJa d nr l, welphgreewaf.fue/ N.Ar a ave *pv The following design bases for the spent fuel and new fuel l.

storage facilities are discussed in the sections indicated below: ,

l'

a. Seismic and system quality group classifications - l i Sections 3.2.1 and 3.2.2

• 9.1-0 Amendment 3 I

, s

's

!

• S ,

Attachment VI (cont'd) l HCGS FSAR 10/83 9.1.2.2.2.2 High Density Spent Fuel Storage Racks

~

High density spent fuel storage racks in the fuel pool store spent fuel, transferred from the reactor vessel. These are top-entry racks.

The spent fuel storage racks are of freestanding design and are not attached to either the fuel pool wall or the fuel pool liner l

' plate. The racks are constructed of stainless steel, and the tc: *~ ruel 1 *-W i:::::: :: :::h ::::te::tia.. . :: r!2 2 neuttM gbwrker is Qotal. Set Fi ute : N:l'3 Gr - : ^'%?h delnift tuk. d'H"

$g fp{ neck anA fhe S .

4 e

t s

P i

9.1-10b Amendment 2

. Attachment VI (cont'd)

BCGS FSAR

. 39 l The spent fuel pool has been designed !for a storage capacity for i 40F4 -t Ir:rt !!!!' fuel assemblies, plus J h' multipurpose cavities for A storage of. control rods, control rod guide tubes, and defective fe containers.. For initial plant 1078 o 15::: ?cee* fuel assembliesgrill;gperation,

_. Provideda and storage space the remaining )(

storage capacity will be added later.

(Yfin 30 mulHparpse cavihir)

Ti e.g...i. si.c; cf the i;;ct:r i;ildin'7terves == n low .

Geakage barrier to provide atmosnheric isolation Et' gie spent I an asp ciated fuez nandling a'reas fuelsporag, us~ + n a ,, = a y = r-: s w a am 9.1.2.2.2.3 Refueling Area Cavities l

As shown'on Figure 9.1-2, the cask loading pit and the reactor well are adjacent to the spent fuel pool. Tne dryer and separator pool is adjacent to the reactor well. Like the spent fuel pool, these cavities are lined with stainless steel plate and are provided with liner leakage collection systems. The reactor well and the cask loading pit are connected to the spent fuel pool by fuel transfer canals approximately 4-feet wide.

Each canal is provided with two gates and concrete plugs to prevent loss of water from the spent fuel pool during periods

/ when the adjacent. cavity is not filled with water.

The cask loading pit is designed to permit the underwater loading of spent fuel assemblies into spent fuel shipping casks. The pit can be drained of water during periods when cask loading operations are not being performed. The spent fuel shipping cask can be decontaminated either in the cask loading pit or in the cask washdown area on the refueling floor adja. cent to the cask loading pit.

The reactor well is a circular cavity located directly above the primary containment. Removal of the drywell head and reactor vessel head provides direct access from the reactor well to the inside of the reactor vessel. The reactor well is filled with water during transfer of fuel assemblies from the reactor vessel ,

to the spent fuel pool. Seals are provided at the bottom of the reactor well between the drywell and reactor well wall and between the reactor vessel and the drywell to prevent water

~

leakage.

The dryer and separator pool provides for storage of the steam dryer and steam separator when they are removed from the reactor vessel. The dryer and separator pool is connected to the reactor 9.1-11 t

Attachment VI (cont'd)

RCGS FSAR 4/84 ,

i well to permit underwater transfer of components between the two cavities. Concrete seal plugs are provided to minimize water loss during normal and abnormal storage conditions when the reactor well is not filled with water. Gaskets are attached to the hori.gontal surfaces of the seal plugs to further reduce water loss.

l 9.1.2.2.2.4 Other Features

- l The spent fuel pool area ventilation system is discussed in l Sectior. 9.4.2.

The area radiation and airborne radioactivity monitoring instrumentation is described in Section 12.3.4.

l l

9.1.2.3 Safety Evaluation 9.1.2.3.1 Criticality Control ,

Geometrically safe configurations of fuel stored in the spent fuel array, and poison materials, are employed to ensure that K,gg will not exceed 0.95 under any normal or abnormal storage condition. To ensure that the design criteria are met, the following normal and abnormal spent fuel storage conditions Mg' analyzed: " '8

a. Normal positioning of fuel assemblies in the spent fuel storage array

b. Eccentric positioning of fuel assemblies in the spent fuel storage array

c. Normal storage array of ruptured fuel j l

d. Moving or placing a fuel bundle along the outside of i storage racks 2

e.. Deleted l t

9.1-12 Amendment 5

(cont'd)

Attachment VI

- HCGS FSAR 6/84

f. Spent fuel bundle falling onto the rack with spent fuel l

g. 2.--

_ De"lA'fe5E;l rre' f:llin; : t:

t' ree' rit5 :;;;t it;;.

(* AV 9.1.2.3.2 High Density Spent Fuel Rack Design Criteria The principal design criteria of the spent fuel racks are as follows:

~

a. QIerrt

-. @BQ fo ?f_. fuel assemblies may be stored in the fuel pool.

b. The storage racks provide an individual storage compartment for each fuel assembly. The. fuel assemblies are stored in a vertical position with the lower tie plate engaged in a captive slot in the lower fuel rack support plate.

c. The weight'of the fuel assembly is held by the lower rack support plate.

d. The spent fuel storage racks are made from 304L stainless steel.

e. The minimum center-to-center spacing for the fuel J'-

assembly L;tu;;;-::x;7 nf the =irirr criter +a-r :t;.

W within the rows ;i!! 5: ; :cil;J yci; te f :14-4eadt- Fuel assembly pl ment between rows is not possible. ' s k wst a fi y ef.l-3, 8--

f. Errf-in tro 1 23 ::t sold.; et th; tr; ;' ;L. 6.ck;

((0[ff",3"ki""t?_5bII'iI"'I222;"I I S'#1"i IC22 IiO

~YOlefet(

g. The impact force considered in the rack design will be provided prior to fuel load. -

h. The storage rack is designed to withstand a pull-up force of 4000 pounds and a horizontal force of 1000 pounds. There are no readily available forces in excess of 1000 pounds. The rrrh; ;;; f i;::e rith '-

1,.J' .i; i: ;;;;;;t ;ti klag. S r;;. 3 inp- the event J of a stuck fuel assembly, the maximum lifD ng force kn "

9.1-13 Amendment 6

. - - _.1

-e 56 8 0 6- - 6' - c3 5 - . N .' o N~ v - v ~

Attachment VI (cont'd) d

, [/ BCGS FSAR i

s e.;--- af inaa ;~2-fr. "h: ::h  ::: f:-i- rf ritM '- #

12:2 e.te te p;:rert stiching. ....e., in the e crt ,

f e .i.sk !.;l ee.=-i-1, 1 iLe

:;ir : liftin; f;;.. of the fuel handling platform grapple, assuming limit i l

switches fail, is 3000 pounds.

l

i. The maximum stress in the fully loaded rack in a faulted condition will be provided prior to fuel load.

j. The spent fuel storage racks also have the capability of storing control rod guide tubes, control rods, and i defective fuel containers. When the spent fuel is i' stored in the spaces provided for storing the above the Kegg does not exceed 0.95.

k. Several design features reduce the possibility of heavy l l objects dropping into the fuel pool. The main and i auxiliary hoists of the reactor building polar crane i are single-failure proof. In addition, the main hoist i

, is physically prevented from traveling in the truncated  !

segment shown on Figure 9.1-31 by mechanical stops on the girders of the polar crane. The~ crane design is discussed in Section 9.1.5. The removable guardrail and the four-inch curb around the refueling cavities further limit the possibility of heavy objects dropping <

into the fuel pool.

1. The fuel storage pool has water shielding for the i stored spent fuel. Liquid level sensors are installed to detect a low pool water level. Makeup water is ,

available to ensure that the fuel will not be uncovered ,

should a leak occur. j i

m. Since the fuel racks are made of noncombustible 1 saterial and are stored underwater, there is no potential fire hazard. The large water volume also e protects the spent fuel storage racks from potential pipe breaks and associated jet impingement loads.

9.1.2.4 Soent Fuel Rack Inservice Insoection An inservice inspection program is in effect throu.ghout the life of the racks to ensure that the quality of the poisoned racks is 9.1-14 i

u i l i m a Mwi E** Attachment VI (cont ' d)

@O,,*"'" M' j -

e e

so.veet us. c oit:4.o,3n e =

m.665 u se II888N 8'E8"We l T YP.

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• M** HOPE CREEK

+ + 4 /,9 + A GENERATING STATION O + + f4 d o FINAL SATfiTY ANALYS18 REPORT

-3 .n d j

~ ^

801 TOM VitW - T-mu i o c gg *y gh FIGURE 9.14; i

i t

?- - - .. ._._. .. ._. __,_. , __._,_

Attachment VI (cont'd)

Y.

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16 h: <

I ,

a .

i  ; -

l h - s.g m l I , + + [

l sum set. /

l

~,{ t = st "

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\- 69 J

p 4;l:,,-

I evif1P. ele F.

N.%. -

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4

, ;,,7 _

tpt C T IOed E-3 HOPE CREEK -

GENERATING STATION l FINAL SAFETY ANALYSIS REPORT A TYPICAL SPENT FUEL RACK FIGURE 9.1-3 $keef J pCy

1 j l [ Attachment VI (cont ' d) k

0M 2 IS).id MAW.

nu m.so nin. o

,  ;: "j iso.as nxx.

e

"" ns.ts nin.

I I I i 3l t i 1%

r tttV. 1"Ib. 29

,, = -, y .IS Ibb

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f

" f

) ELEv. 15"I .*IS MEUTROM ABSORBER j I- l l

I 3 th .Ib

[ l l l 1f i l l 1 h I '

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)f  :

p / , F

, a NEUT RON ABSORBER e

ELEV. 12.75 i l i i ELEV. B.1%

i

! E LEY . B .%Q j .1% E % '

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- ELEV. t.oo

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(1.0* OF AOJos1 MENT) .

l-

+- -- b . 0 0 r HOPE CREEK D'A- GENERATING STATION

" " ^ " ' " ^ " ' ' ' " ' "

SECTION A-A SHOWN B MIN. HEIGHT A TYPICAL SPENT FUEL RACK i

j FIOURE 9.13 kghaf Y 4

i

Attachment VI (cont'd)

(

12 u 14 -

EHOWN .

t 2 S$. bl 2 i

l l ,

l i ,

I*

1 180.25 t%u.

nus w.

/ v r p -

/ ~~

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+ 4 + 4

• 4' j i + +', + + + e -- ,

, + @ , 4 + + 4 + + + 4 +'

(!,+

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• ,4 4 + + + 4- + +]*__ 4 i + + b 4 + + +

+ & 4 + + + 4 4 +4 "Ib .O + + 4 -a. 4 4- 4 4 4 + 4 , + 4

+ 4

+l4- + + 4- + 4 A 4 + +

4 4 4- '+ 4 4- % & + 4 4 4 9 -+ -+

+ l+ 4 + 4- + + +- + +

+ + F 4 5 4 4 4 -- 4- i* + + 4 - + 4

+ G * - e '*-l+ ++ + + + O!4 i

y + + * * + + - + t + t * +l4 1 4

~

r SS . b*l BOTTOM VitW i _

HOPE CREEK SENERATING ST#. TION r' FINAL SAFETY ANALYSIS REPORT

. A TYPICAL SPENT FUEL RACK FIGURE 9.13 gQf l,C 4

, . . . , , , ,,_,,,,_n. . _ , , , , . , . - , . - . _ - _ _ . _ , , , - -

ccus tsan

- - ~

~~ was Attachment VII *

(-

• Dv w roN 410.42 (SECTION 9.1.2) l Specify the weight percent of D535 which corresponds to he

• hig' hest anticipated enrichment used in the criticality calculations for the new and spent fuel storage areas.

l l

l The highest anticipated assembly. average enrichment of U2s5 used*

2.f&&

in the spent fuel rack criticality analysis -ill i ;cesilat,1; by Ex:, ? ? ? 'T" _ _ - . . - JJ a w 3 C ie. 0 .1. L" ~,

'The new fuel stocage area criticality analysis is based orr a.

maximum anticipated reactivity of the fuel and not on enrichment alone. The use of reactivity takes into account the combined effects of enrichment, enrichment distribution, gadolinia, gadolinia distribution, and fuel lattice geometry (i.e., water rods and fuel rod pitch) The BCGS new-fuel storage racks were designed to store fuel which has an infinite neutron multiplication factor of 11.21, irt the uncontrolled. reactor core geometry. .

This fuel reactivity limit bounds all existing and expected GE

. fuel designs. The use of the maximust reactivity of the fuel,

t. instead of enrichment alone, is consistent with the NRC Standard Review Plan 9.1.1, as well as industry standards.such as ANSI

,,.. Standards N208, R209, and N110. .

i

\ -

i i

( .

410.42-t Amendment 2. ',

.Y.b n ... : . .

= ~~^

~w+;g

%:15 M CI CZfyl W fif y Dj y W H CGS FSAR ~

,t 282&edbgg'

'V:.~ N.~k 4/84 "I"Nyo*/ryd Attachdent VIII' n .c. -. ' th ; '*~ -

• ~ "'*Te )

Seismic qualification for this isolation system is in /

accordance with qualification procedures and acceptance criteria defined in IEEE Standard 344-1975, and implemented by Regulatory Guide 1.100, Revision 1.

- This isolation system is located in and qualified for a mild environment as defined in Sections 3.11.2.4 and 3.11.2.5.

~

The worst-case specified environmental conditions in which

.this isolation system is designed to operate are as follows:

Pressure: Atmospheric plus fractional inch of H,0 Temperature: 1040F maximum 1 these conditions may 400F minimum exist 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per year 83*F 2 2*F l Relative Humidity: 50% maximum (summertime) 20% minimum (wintertime)

Nuclear Radiation: 175 Rads Carbon (40 year TID) 88 Rads Carbon - Beta (180 day TID) 2.5 Rads Carbon - Gamma (180 day TID)

(TID = Total Integrated Dose)

~ %5vWl Testing 41 n accordance with SAMA Standard PMC 33.1-1978 rill 4- I 5 '9'd, to ensure 9that this isolation h sy:s::m:p.eted te is adequately by Om.:, protected against the effects of electromagne1;ic interference (EMI).

-per % Ned Testing in accordance with JEEE Standard .472-1974 will i;4L k ec=pleted by Pa==, ??"O,' M ensur that this isolation x system is adequately protected ag(inst a the effects of short-circuit failures, voltage faults and/or surges.

b. Computer Products Inc. (CPI) Emergency Response Facilities

~

Data Acquisition System (ERFDAS) - this system utilizes the CPI real time peripheral (RTP) system for 1E to non-1E isolation. The basic components of the RTP system are analog and digital surge cards (qualified to IEEE Standard 472-1974 requirements), analog input cards and optically isolated digital input cards, distributed input / output controllers (DIOC) and transformer-coupled multi-drop .

limited distance modems (MDLDM). The MDLDMs provided the 1E to non-1E isolation. Data transmission to receiving MDLDMs is by twisted-shielded pairs.

~

Seismic qualification for this isolation system is in accordance with qualification procedures and acceptance criteria defined in IEEE Standard 344-1975, and implemented by Regulatory Guide 1.100, Revision 1. I 421.13-2 Amendment 5 s c4% w ~- . .w , .

.c

\

HCGS FSAR 4/84

. Attachment VIII (cont ' d)

This isolation system is located in and qualified for a mild environment as defined in Sections 3.11.2.4 and 3.11.2.5.

The worst-case specified environmental conditions in which this isolation system is designed to operate are as follows:

Pressure: Atmospheric plus fractional inch of B,0 Temperature: 1040F maximum ) these conditions may 40*F minimum j exist 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per year 83*F t 20F Relative Humidity: 50% maximum (summertime) l 20% minimum (wintertime)

Nuclear Radiation: 175 Rads Carbon (40 year TID) 88 Rads Carbon - Beta (180 day TID) 2.5 Rads Carbon - Gamma (180 day TID)

FN*

Testing in accordance with SAMA Standard PMC 33.1-1978 till  %

-be campleted by 1";rrt, 1 9 8 6 ,- 2t : ensurgfthat this isolation k system is adequately protected against the effects of electromagnetic interference (EMI).

i ~pu 41Mut in accordance with IEEE Standard 472-1974j r!!! br E Testing X cc pleted by F^5rrery, '9"' Abd' ensuref that this isolation X

( system is adequately protected against the effects of short-circuit failures, voltage faults and/or surges. These tests dtC eill on performed on the analog and digital surge cards and the transmit / receive circuits of the MDLDMs.

c. Technology for Energy Corporation (TEC) Radiation Monitoring l System (RMS) - this system utilizes three separate isolation methods depending upon the type of isolation required:

1) 1E to IE isolation - for this type of isolation, Hewlett Packard NFBR 1000 and HFBR 2001 isolators are used. Optical coupling is used to provide the isolation.

2) 1E to non-1E annunciator outputs - for this type of ,

isolation, Agastat Model EGP isolation relays are used. l Relay coil to contact separation provides the i isolation.

l

3) IE to non-1E communication - for data transmission between the TEC IE microprocessor and the non-1E host computer, TEC Synchronous Data Link Control, serial communications modules 600-1200 are used. Transformer coupling provides the isolation for the transmit circuits. Optical coupling provides the isolation for the receive circuits.

I 421.13-3 Amendment 5 l

--_-m

1 HCGS FSAR 4/84

~

L Attachment VIII (cont'd)

Seismic qualifaction for these isolation systems is in )

l accordance with qualification procedures and acceptance criteria defined in IEEE Standard 344-1975, and implemented by Regulatory Guide 1.100, Revision 1.

l These isolation systems are located in and qualified for a l

mild environment as defined in Sections 3.11.2.4 and l 3.11.2.5. The worst-case specified environmental conditions in which these isolation systems are designed to operate are as follows:

Temperature: 1040F maximum these conditions may 400F minimum exist 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per year 760F t 20f Relative Humidity: 50% maximum i

-n se 'isolohon 20% minimum arc

~ fthA Testing in conformance with Military Standards 461B and 462 shstenes e

9g,(

[*ka;ttfu on the ffe fEMI{cillbe:::pletedbyJuly, 1994.8- g e.ds o( EMI ,

TestingfiaccordancewithIEEEStandard 472-1974. ull W e

plet_d by July, '99f, t

4ensurejthat these isolation systems are adequately protected against the effects of l short-circuit failures, voltage faults and/or surges. .

d. Remote control panels - two isolation methods are provided for remote control panels requiring IE to non-1E isolation.

1) Digital IE to non-1E isolation - for this type of isolation, Struthers Dunn type 219, Allen Bradley model 700-200A12P, and General Electric model HEA99 isolation relays are used. Relay coil to contact separation provides the isolation.

2) Analog 1E to non-1E isolation - for this type of isolation, TEC analog isolators, model 156, are used.

Transformer coupling is used to provide the isolation.

Seismic qualification for these isolation systems is in accordance with qualification procedures and acceptance criteria defined in IEEE Standard 344-1975, and implemented by Regulatory Guide 1.100, Revision 1.

The Struthers Dunn type 219 and General Electric model HEA99 irolation relays are located in and qualified for a mild environment as defined in Sections 3.11.2.4 and 3.11.2.5.

The worst-case specified environmental conditions in which these isolation relays are designed to operate are as follows:

Struther Dunn Type 219 i

4 421.13-4 Amendment 5 9 "l i

. - _ _ - . _ . _ - . . - - _ . _ _ _ _ _ _ . _ _ . . _ _ . _ . . . . . . , . - _ . - - - . _ - . _ _ . , . _ _ . _ , , _ . . . ~ . . . _ . -

~ ' ~

- HCGS FSAR 4/84 Attachment VIII (cont ' d) l Nuclear radiation: 200 Rads (40 year TID)

No testing was conducted on the effects of EMI on the l Struthers Dunn type 219, Allen Bradley model 700-200A12P, or l General Electric model HEA99 isolation relays. By design, l these relays should be immune to the effects of EMI.

Ger etic EMI susceptibility and emissions test were conducted on'the TEC model 156 analog isolators following procedure 156-OP-04, " Electromagnetic Interference (EMI) Test for TEC Model 156 Analog Signal Isolator Module," which is Appendix B to test report 31041-OP-01, " Qualification Test Report for Environmental and Seismic Testing of the TEC Model 158 Analog Isolation System." Results of these tests are available for review at Technology for Energy ,

Knoxville, Tennessee. j Corporation,ke*4d

- pec Testing /inaccordancewithIEEEStandard 472-197(/ r!!! i;4-p f;:::d i:'ensurelthat the,Struthers Dunn Type 219 ende Geuws .1 Elvs m ;; ::fr: 22J.;, isolation relays are adequately protected against the effects of short-circuit failures, voltage faults and/or surges.

-/NSE2T~B -

The following test was performed on the Allen Bradley model .

700P-200A12P isolation relay to ensure adequate protection ,

against the effects of short circuit failures, voltage faults and/or surges:

1) Test type - 100% high potential test

2) Test characteristics - 2700 V applied for one second between points of opposite polarity and to ground.

,.--park med Testing;in accordance with IEEE Standard 472-1974f rill 9 O*h"Y) 7 rfer :f i ensurefghe TEC model 156 analog isolators are 8 \_

adequately protected against the effects of short circuit failures, voltage faults and/or surges.

o. Equipment air lock isolation dampers HD-9450A and B interlock with receiving bay door 44323A - Potter Brumfield model MDR isolation relays are utilized to provide both non-IE to IE and IE to non-1E isolation as shown below:

1) Non-1E to 1E - receiving bay door 44323A (non-1E coil) permissive to equipment air lock isolation dampers HD-9450A and B (1E contact)

2) 1E to non-1E - equipment air lock isolation dampers HD-9450A and B (IE coil) permissive to receiving bay door 44323A (non-1E contact)

These two relays were purchased from General Electric.

421.13-6 Amendment 5

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- HCGS FSAR #

4/84 Attachment VIII (cont ' d)

f. Startup Transient' Monitoring System (STMS) - The qualification requirements of isolation devices, used by the STMS are described in Section 7.5.1.3.5.

,gg, - /AIS&/E'7~ A -

$The isolation devices used to electrically separate nonessential and essential circuits are pursuant to the guidelines of IEEE Standard 384. Both relay and optical isolation devices are employed. The optical isolators utilize.a fiber-optic light pipe to electrically separate the input from the output. For example, an essential logic signal activates a light emitting diode, the light is transmitted through the light pipe to a photo switch and the switch changes state on receipt of the light signal and either blocks or transmits.

The relay isolation devices provide the same degree of separation and are used typically for control voltage separation applications, i.e., 120-Vac and 125 Vdc essential to nonessential and redundant essential circuits. The relays are mounted so that a metal barrier separates the coil from the contacts with a minimum distance of one inch between the coil and barrier and

between the contact and barrier.

Summary of Purchase Specification:

' a. RELAY b. ISOLATOR

1. Design Specification 1. Bill of Material a) MIL-R-19523 b) Contact Specification c) Coil Specification d) Insulation Specification e) Design Life f) Reliability l

2. Class 1E Safety Function 2. Purchase part )

drawings 204B6186 and 204B6188

! a) Functional Specification i b) Reliability

3. Qualification Testing 3. Qualification Testing i a) Ambient and Design a) Tested as i Environments a panel b) Application Configuration

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. . Attachment IX -

,3 HCGS FSAR 4/84

' 4 I

'each individual will meet the education and experience h- requirements of ANS/ ANSI 3.1 - 1981 prior to initiate fuel loading.

1 l

I In general, the personnel assigned to the licensed operator l

training come from one of the following areas:

1. Degreed engineer l 2. Previously licensed (BWR/PWR)

} 3. Navy nuclear plant operator

4. Fossil plant operator

5. Sal.em EO upgrade In general, personnel assigned to the non-licensed operator i training will come from one of the fol1~owing areas:  !

1. Qualified utility / equipment operator from Salem i Generating Station

2. Navy nuclear plant operator pog g/ 3. Fossil plant operator

b. Training on the HCGS plant specific procedures and technical specifications will be conducted as the procedures become available. These procedures are under development and will become available at various intervals throughout the training period. To ensure that all licensed operator f )

andidates are thoroughly _ familiar with the procedures and

.A, J odA M, echnical specifications,Aan intense pre-11 cense training 33ert 8  !

1. program]will be implemented three (3) to six (6) months .

prior to the license examinations. This training will cover l all the HCGS specific operating, abnormal and emergency j procedures, administrative and emergency response procedures, technical specifications and low power and surveillance testing procedures. Training will be covered by classroom instruction, in-plant oral examinations, written examinations and performance testing on the Hope Creek specific simulator.

i I

i

c. Applicable references for each of the segments outlined in the appendices are shown on the appropriate cover sheet of each appendix.  ;

d.

Training segments which include 10CFR Part 55 Section 21, 22 and 13G.

23 are identified in Appendix 13A, 13C, 13E, 13F and 1

e. The following segments of the training program are still under development: .

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kid Iocense opero4cr I

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ik kh \S i.) Mk N h odh 4e % neq . . O. i O l se: ' /dNMb[' ^* -f_'_3]);QJ 4b ~ _. -- = (cont'd) . ',, Attachment IX 4/84 . HCGS FSAR g O f. x====;=4=',: 11 the cc-tent - u_ _.. l._m ' rr' rr; crt -.11 c - gr===e=r== d;==rieir;; w L,m Jm.mioped my emum .,o. 4 ,

g. Hot license training for NRC candidates will be conducted to augment the shift staffing allotment, allow for promotion or fill vacancies due to reassignment. This training will utilize a major portion of the existing cold license training program; however, certain areas may be waived based on an individual's prior experience and educational background. Procedures describing the content and I administrative requirements will be completed by June 1985.

h. Appendix 13F has been revised to incorporate this response. l W .c'A 6tH8t A cost se de.scr;pkn for gegenen+ s [ and } o ( \ he is ec>n+ained i n A ffendltes

'trainin3 p rogram j i 3 s' a n ol n s j , r es pe c+i tle I . t i 630.7-3 Amendment 5 + + y ATTACHMENT'X' PAGE l OF 2-HOPE CREEK GENERATINGcSTATION FSAR COMMITMENT. STATUS.THROUGHEJULY 1984 s PSAR COMMITMENT LOCATION ' COMMITMENT RESOLUTION l1. .Ouestion/ Response: This commitment concerns 1provid-Appendix:- _ _ ing/ design data ito demonstrate Question 210.12. compliance.with Paragraph .. . NC/ND-3649 of: Code Case:N-192. This inf ormation has been pro-vided . in the letter, - R. .L. Mittl (PSEG)-to A. Schwencer (NRC), " Compliance withl Reg. Guide' l.84", dated July 3 0 ,. '1 9 8 4 . The information in Attachment XI will be included in Amendment.8 to the HCGS-PSAR. 2.- 10uestion/ Response- This commitment concerns provid-Appendix:- ing the method. and acceptance' Ouestion 210.20 criteria for a dynamic analysis of' the f eedwater check valve re-sponse to a feodwater line break outside containment. This infor-mation will be provided in August 1984.

3. Question / Response This commitment concerns prepar-Appendix: ing preventive maintenance proce-Question 410.87 dures for instrument air systems in compliance to ANSI MC11.1-1976. This information will be provided in November 1984. This revised commitment date will be included in Amendment 8 to the HCGS FSAR.

4. Question / Response -This commitment concerns itemiz--

Appendix: ing results of a review to de-Question 210.12 .termine if bypasses may have to be used for in-service testing of . Electric Power and Protection Systems. This information has been provided in the letter; R. L. Mittl (PSEG) to A. Schwencer (NRC), "DSER Open Item Status," dated August 3, 1984. M P84 123/13 .1-gs - . s PAGE 2 OF 2 ~ . FSAR COMMITMENT LOCATION COMMITMENT RESOLUTION 5.- Question / Response This commitment. concerns testing Appendix: . the isolation systems . for the -Ouestion 421.13c Radiation Monitoring System for effectsHof EMI-and protection against effects of short-circuit failures, voltage faults, and/or . surges. This information has been provided in the . letter; R. L. Mittl (PSEG) to A. Schwencer (NRC), "DSER Open Item Status", dated August 1, 1984.

6. Question / Response This commitment concerns provid-Appendix: ing information on the capability Question 421.22 for the at-power surveillance testing of the instrumentation channels, logic and actuation de-vices of plant safety systems.

This information has been pro-vided in the letter; R. L. Mittl (PSEG) to A. Schwencer (NRC), "DSER Open Item Status," dated August 1, 1984.

7. Question / Response This commitment concerns provid-Appendix: ing justification of BWR reactor Question'421.23 vessel level sensing lines. This information has been provided in the le t ter; R. L. Mittl (PSEG) to A. Schwencer (NRC), "DSER Open Item Status," dated August 1, 1984.

8. Question / Response This commitment concerns prepar-Appendix: ing an emergency load sequencer Question'430.19 (ELS) system reliability analysis. This information has been provided in the letter; R. L. Mittl (PSEG) to A.

Schwencer (NRC), "DSER Open Item Status," dated August 1, 1984. The information in Attachment XII will be included in Amendment 8 to the HCGS FSAR.

9. Question / Response This commitment concerns provid-Appendix: ing license examination periods Question 630.9 in FSAR Figure 13.2-1. This in-formation has been provided in Amendment 5 to the HCGS FSAR.

M P84 123/13 2-gs l Attachment XI l HCGS FSAR 1/84 and/or1.84. Guide 1818 subject to the limitations recommended by Regulatory Code Case N-192 was invoked in the fabrication of certain . flexible metal instrument hose assemblies and on certain standby diesel generator skid-to-facility connectors. Regulatory Guide 1.84 states that this code case is acceptable subject to the requirement'that the applicant should provide design data to demonstrate compliance with paragraph NC/ND-3649. Informa' tion to comply with this additionql regulatory requirement u;11 L= gee.idrd ir Jun: !?St. Code Case N-275 was invoked in the fabrication of certain safety-related pipe. Regulatory Guide 1.84 states that the design guidance in this code case is acceptable subject to the additional welding restrictions in the regulatory guide. HCGS complies with these additional regulatory requirements. The HCGS piping design specification permits the use of Code Case N-275 subject Guide 1.84. to the limitations recommended by Regulatory Code numerous Case 1644 and its various revisions has been invoked in applications. Regulatory Guide 1.85 states that this code case is acceptable subject to the limitations on maximum ultimate tensile strength and, in the case of Code Case 1644-9 (N-71-9), the additional requirements for electrode dispersal. HCGS is currently evaluating the applicability of the additional maximum ultimate strength limitation in view of the concerns with material will be providedbrittleness in June and1984. stress corrosion cracking. A response Use of Code Case 1644-9 (N-71-9) is subject to the additional precautions cited in Regulatory Guide 1.85. Use recombiner of Codetechnical Case N-249 is permitted for the containment hydrogen specification. To date, this code case has not been invoked. Code Case N-253-1 provides rules for the construction of ASME components which experience elevated temperatures. This code case was invoked in the design of the containment hydrogen recombiners.

• This code case was invoked on HCGS because there are at temperatures portions of the in excess containmentof 8000F. hydrogen recombiners that opera h -

9 Q fegn rafsm;$N undet ci,oses)C Cover Y *$ .. . (f6Eb6h LO A W?^ 5 A)Mf aba/td J0lf ' SN') 198 h. SeCGst .Ng_ jsyg,,nq/,b n } .5 Cons.*elertal /s je na:/h 4a /d ke IQh'nden hom 'yapr19}s q reg.sgS/ /0 FU b Nc. e6:3c/e.$vr4 ib MClNfd In /Ja. /4/M. 210.12-2 Amendment. 4 . - . . , - , , - . , . - - _ _ _ ., .m--__ _ , . , , , _ _ _ . - . . _ , , , . _ , , _ , , , _ -.-..,.....-____,_..........-,,-,_m.-- , - . , . , t * . Attachment XII ~ i BCGS FSAR g +

f. LOP AFTER LOCA swupCING COMPLETED If a LOCA signal is still present when the SDG circuit breaker is cGosed, the LOCA signal overrides the LOP sequencer and starts the LOCA sequencer to apply LOCA loads  !

in the predetermined sequence. , For scenarios '2a' through '2f' above, the PSIS signals are present to prevent the inadvertant starting of equipment before its predetermined sequenced time. ELS TESTING: Provisions exist at each of the sequencer cabinets to test the ELSs for 2a through 2f scenarios described above. An alars is provided in the main control room to indicate that an ELS is being tested. If an. actual LOP or LOCA occurs during the testing of an ELS, the sequencer resets autokatically and responds to LOP and/or LOCA event. f_ / -> 1 The ELS system reliability analysis * ^ - - - ' " * " *

• u M DE & A j SEthSATE C.o v 6 1 . THE ELS SWTEMEE LIAW Lt TY 18 EW mMco BY Tm u se of TWO REmuNb44T Micto Pto-CESSOES N EACH oF TWE. TOVE. ELS, SYSTEMS.

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"bseg. Op 1% W' "j dded %"s*I 's '424) i i . 430.19-3 nt / . -__-_.-.:---....- _ - _ - . - - - . _ _ . - - . - - . . . -