ML20090L108

Forwards FSAR Commitment Status Through Apr 1984,per Request for Addl Info & Responses to Questions 251.1,220.16,430.94 & 640.9

ML20090L108
Person / Time
Site:	Hope Creek
Issue date:	05/16/1984
From:	Mittl R Public Service Enterprise Group
To:	Schwencer A Office of Nuclear Reactor Regulation
References
MO-16-01-02-D, MO-16-1-2-D, NUDOCS 8405250045
Download: ML20090L108 (21)

Text

.

Pubhc Sorvce O PS G Comr.any Electnc and Gas 80 Park P! ara, Newark, NJ 07101/ 201430-8217 M AILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation May 16, 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 PSAR COMMITMENT STATUS THROUGH APRIL 1984 Public Service Electric and Gas Company presently does not plan to issue Amendment No. 6 to the Hope Creek Generating Station Final Safety Analysis Report before July 1984.

Accordingly, this letter is provided to document the status of Hope Creek Generating Station responses to NRC requests for additional information which were forecasted to be responded to by April 1984.

Attachment I ic a tabulation of the Hope Creek Generating Station Final Safety Analysis-Report commitments for April 1984 and the corresponding resolution for each commitment.

Attachments II through V provide.the responses to the questions forecasted to be responded to in April 1984, which will be included in Amendment No. 6.

Public Service Electric and Gas Company submitted Amendment No. 5_to the Hope. Creek Generating' Station Final Safety Analysis. Report on April 30, 1984 (R. L. Mittl, PSE&G to A. Schwencer, NRC). Amendment No. 5 incorporated the responcen to prior NRC requests _ for additional information forecasted for_ closure-in February and March 1984, as I

outlined in PSE&G^1etters of March 5, 1984,'and April 2,

-1984, (R. L. Mittl, PSE&G to A. Schwencer, NRC). O

~ :p The Energy People 84j5 g g5 j4 g g4 \

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, Director.lof Nuclear 2 5/16/8d Reactor' Regulation Should you have any questions in this. regard, please contact us.

Very truly-yours, bi h Attachment I - Hope Creek Generating Station - FSAR Commitment Status through April 1984.

Attachment II - Response to Question 251.1 Attachment III - Response to Question 220.16 Attachment IV - Response to Question 430.94 Attachment V - Response.to Question 640.9 C D. H. Wagner (w/ attach)

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

USNRC Senior Resident Inspector:

MO 16'01/02-D l'

Page 1 of 2 ATTACHMENT I HOPE CREEK GENERATING STATION FSAR COMMITMENT STATUS THROUGH APRIL 1984 FSAR Commitment Location Commitment Resolution

1. Question / Response This commitment concerns demonstration Appendix: of compliance to 10CFR50.55a and Question 251.1 Appendices G and H of 10CFR Part 50 for ferritic reactor coolant pressure boundary materials. The response:tx) this question is provided in Attachment II and will be included in Amendment 6'to the HCGS FSAR.

2. Question / Response This commitment concerns a description Appendix: of the spent fuel racks conformance Question 220.16 with applicable provisions of sub-section NF of the ASME-Code. 'The response to this question is provided in Attachment III and will be included in Amendment 6 to the HCGS FSAR.

3. Question / Response This commitment concerns submittal of Appendix: a comparison of response spectra Question 220.21 results from the finite-element and half-space methods. .This information will be provided in July 1984.

4. Qaestion/ Response This commitment concerns ' submittal of Appendix: a description of diesel' generator Question 430.94 fuel oil delivery to the site during flood conditions Jand the procedures used for refilling the storage tanks during flood conditions and non-flood conditions. The response to this question is provided in

Attachment:

IV

- and will be included in Amendment 6 tx) the HCGS FSAR.

5. . Question / Response 'This commitment concerns submittal _of Appendix: information.'on laboratory tests Question'460.4 conducted'under the process; control program (PCP)'for the solidification of solid radwaste, and'on the.

. laboratory / field _ instruction' record sheet:to be used within'the:PCP.

This information will'be provided in January 1985.

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. i Page 2 of 2 FSAR Commitment Location Commitment Resolution

6. Question / Response These commitments concern submittal Appendix: of information on training programs Questions 630.7 for licensed and non-licensed opera-630.9 tions personnel, schedules for 630.10 examinations prior to fuel loading 630.12 and after criticality, training programs for management personnel and technical support staff, and fire protection training. This information has-been provided in Amendment 5 to the HCGS FSAR.

7. Question / Response This commitment concerns submittal of Appendix: the results of our findings on Question 640.9 adequacy of the drainage in affected areas to preclude flooding, upon automatic sprinkler actuation. The response to this question is provided in Attachment V and will be included in Amendment 6 to the HCGS FSAR.

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1 Attachment II HOPE CREEK QUESTION 251.1 Appendices G and H,10CFR50 were revised in the F6deral Register on May 3

27, 1983 and became effective on July 26, 1983.

QUESTION 251.la i

i Identify ferritic reactor coolant pressure boundary materials that do not comply with the fracture toughness requirements of Section 50.55a and Appendices G and H of 10CFR Part 50.

RESPONSE

A major condition necessary for full compliance to Appendix G of 10 CFR

! Part 50 is satisfaction of the requirements of the Summer 1972 addenda to Section III of the ASME B&PV Code. This is not possible with the HCGS reactor pressure vessel (RPV) and the main steam isolation valves (MSIVs), which were purchased to earlier B&PV Code requirements. The i RPV was qualified to the applicable General Electric RPV purchase i specifications and by toughness testing in accordance with the 1968

,' edition of Section III of the ASME B&PV Code as well as the addenda through Winter 1969. The MSIVs were built to the draft addenda for Class 1 pumps and valves of the 1968 ASME B&PV Code.

i OUESTION 2511b i

4 For materials which cannot meet the fracture toughness requirements of

} Section 50.55a and Appendices G and H of 10 CFR Part 50, provide

alternative fracture toughness data and analyses to demonstrate their l equivalence' to the requirements of 10 CFR Part 50.

RESPONSE

I For plants that received a construction pemit prior to August 15, 1973, the NRC Branch Technical Position MTElf No. 5-2 provides guidance for 4

making conservative estimates and assumptions that may be used to show ,

j compliance with the intent of the latest requirements. Paragraph 1.3 of i MTEB 5-2 permits the use of other methods that can be shown to be conservative. Based o_n an evaluation of actual toughness' data for j plants of this period and on applicable data from Jeldina Research Council Bulletin 217, General Electric procedure 6I006A006 was derived

.l to estimate- compliance with the intent of Appendix G and to establish t

values for the-initial reference transition temperature (RT of older

!. RPV materials. For other materials, including the MSIVs, mEo)ds for l evaluation of compliance with the intent of current Appendix G requirements are explained' in Appendix SA.-

j -QUESTION 251.1c l To demonstrate conformance to Appendix G and H,,10 CFR Part 50:

! QUESTION 251.1c(1)

Provide pressure-temperature limit curves for hydrostatic pressure and leak test, heat-up, cooldown and core operation.

1

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RESPONSE

See Figure 5.3-1. When Appendix SA was submitted in Amendment 1, this figure was revised to conform to the July 26, 1983 revision of Appendix G.

QUESTION 251.1c(2)

Identify the withdrawal schedule, lead factor, test samples and materials in the Reactor Vessel Materials Surveillance Program.

RESPONSE

The May 27, 1983 revision of Appendix H of 10 CFR 50 requires the withdrawal schedule for the surveillance program capsules to meet the requirements of ASTM standard E 185-82. The lead factors for the HCGS surveillance capsules are 0.86 at the inside surface of the vessel and 1.20 at one-quarter of the way through the vessel wall measured from the inside surface. These lead factors were calculated assuming that the vessel is symmetrical. This assumption was made because the vessel qualification program did not provide for measurements of vessel radii to identify any angular locations where the inside diameter of the vessel is larger than nominal. Hence, it is possible that a surveillance capsule could be located at an extended radius position.

This would provide surveillance sample test results lower than calculated and nonconservative values for the peak fluence when it is estimated from the capsule data using the aforementioned lead factors.

Details of the materials and test samples contained in the HCGS surveillance program are provided in Section 5A.4.

QUESTION 251.lc(3)

Indicate the reference temperature, RT , for materials in the reactor vesselclosureflangeregionandthebhk[lineregions.

RESPONSE

Reference temperature RT flange regions are given N 9[ valves for materials in the RPV closure Section SA.2. Similar values for materials in the beltline region of the HCGS RPV are provided in Tables 5A-4 and SA-5.

QUESTION 251.1c(4)

Indicate the chemical composition (copper, nickel and phosphorus),

unirradiated upper-shelf energy, and projected end-of-life RT and upper-shelf energy for all beltline materials. RT project $DIretobe estimated ~using the "Guthrie Fonnula" in CommissioNDkeport SECY-82--465.

Upper-shelf energy projects are to be estimated using Regulatory Guide 1.99, Rev. 1. These projects are to be for the end-of-life neutron fluence at the 1/4T and ID reactor vessel locations.

RESPONSE

Details of where beltline materials are located in the RPV are provided in Tables SA-4 and 5A-5. Also contained in these tables are the appropriate fluence values used for the shift calculations.

Unirradiated and end-of-11te upper-shelf engines and RT values are giveninTable251.1-1alongvaluesfortheshiftsinRYDT calculated bythe"Guthrieformula"foundinFigureE-1ofCommissiURTReport SECY-82-465. However, the radiation shift values used for the pressure-I temperature limit curves presented in Figure 5.3-1 are those derived from shifts calculated according to the formula given in revision 1 of Regulatory Guide 1.99.

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TABLE 251.1-1 Upper-Shelf Energica cnd *Guthrio" RTHM Y31"33 R.G. 1.99 Rev.1 RTHW( F)'

Projected EOL CHEMISTRY (Wt.%) Initial Upper-Shelf Upper-Shelf "Guthrie" " Gut hr ie*'

Unirradiated Shift EOL Heat 8 or Heat /Plux Cu Ni P Energies (Pt . -Lb. ) Energies ( Ft .-Lb . )

5E2963-1-2 0.07 0.58 0,009 102 89 -10 +48 +58 SE2530-1-2 0 08 0.56 6.010 86 75 +19 +72 +91 SE3238-1-2 0.09 0.63 0.012 76 66 +7 +77 +84 SE3230-1-2 0.07 0.56 a.010 121 105 -10 +6 8 +58 6c35-1-2 6.09 0.54 0.010 107 93 -11 +75 +64 6C45-1-2 0.08 0.57 0.008 97 84 +1 +72 +73 SE3025-1 0.15 0 71 c.012 75 69 +19 +84 +103 SE2600-1 -0.09 6.58 e.012 75 69 +19 +66 +85 SE2698-1 0.10 0.58 0.010 75 69 +19 49 +88 510-01205 0.09 0.59 0.010 >92.5 >80 -40 +76 +36 53040/1125-02205 0.08 0.63 0.010 135 117 -30 +73 43 l 519-01205 0.010 0.53 0.010 >109 >100 -49 +47 -2 504-01205 0.010 0.51 0011 >125 >115 -31 47 +16 55733/1810-02205 S.10 0.68 0.013 >68 >62 -40 +70 +30 53060/1810-02205 0.10 a68 0.012 >95 >87 -49 +70 +21 001-01205 0.02 0.51 0,012 >109 >100 -40 49 +9 19444-1-4,5 0.12 0.81 0.011 >79 >73 -20 +74 +54 10024-1-2,3 0.14 6.84 0.010 ,

>70 >64 -20 +80 +40 (1) This information is for a depth of 1/4 of the wall thickness from the inside surface of the vessel.

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Num. EAR ENEROY BullNESS OPERATIONS GENERALOitiernic nav i

1. SCDPR OF APPLICATIONS AND OBJECTIVR9 1.1 This procedure describes the method to be used for establishing the WI I I'# I#III' '"I initial referosos temperature (RT ****I* I'* 'IS*'

plaats where fracture toughness data may be incomplete. These methods repres at a General Electric alternate positlom to the NRC Rossistion 10CFESO Appendiz G for these plaats.

2. E'IBODS 2.1 Vessel Plate ISA-533 Gr. B C1.1): ,

Predicted limiting property - either NDI (Nil-Destility Transition Tes:perstare) or transysrse CVN (Charpy V-Notch) 50 f t-lb T.T. (Transition Temperature)

Usual data available - ICT and/or longitudinal CYN at +10 or +40*F RTWI predietion method -

i Operate on lowest longitudinal CVN f t-1b to get at least 50 f t-lb T.T. by adding 2*F per it-lb or by plottlag a curve (f t-lb versus temperstare),

where possible. Add additional 30'F to convert from longitudinal to transverse 50 f t-lb T.T.

1 l NOTE: There transverse CVN impact data are available, but the 50 ft-1b T.T. is not met, operate on the lowest CVN f t-1b to get at least 1 50 f t-1b T.T. by adding S*F per it-1b or by plotting a earve

(f t-1b vs temperstars), where possible. This extrapolation is valid for CYN test temperatures only la the range (-25' to +50*F).

Derive NDT, where missias, as egual to longitudinal CYN $$ f t-lb T.T.

1 RT is higher of NUT or transverse.CVN $0 it'-1b T.T. -40*F NUT 2.2 Forlans (SA-50s C1. 2);

Predicted limiting property - WT or transverse CYN $0 f t-lb T.T.

I 4

Usual data available - NDT and/or CYN at slagle temperature RTg.g prediction method -

Derive CVN S0 ft-1b T.T. as for plate.

Then only CVN valses are available, estimate NDT as the lower of +70'F or the CYN test temperstare where at least 100 ft-1b or 30 percent shear is . i schieved. ,

i RT ET is higher of WT or transverse CYN S0 f t-1b T.T. -40*F. I me3 eeFAInew.seteeg l

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, C'UCLEAR INEtCY GENERAL ELECTRIC 0' A"' =. 3 EUSINE55 OPERATl:NS cav 1 FINAL 2.3 Weld Metal (Used to Join SA-533 Gr. B CL.1 Plates and SA-50s CL. 2 Formina s) : . .

Predicted limiting property - CVN 50 f t-lb T.T.

I Usual data available - CVN vaines at single or at several test temperatures.

l RTNDT predicti a method -

Operate on lowest CVN f t-1b to ge t at least 50 f t-1b T.T. by adding 2*F per it-lb or by plotting a curve (f t-lb versus temperature), where

. po sible RT NDT is the CYN 50 f t-lb T.T. - 60*F. If NDT is available, it will be i

considered also. In abse sce of NDT data, RT NDT shall act be lower than

-5 0

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2.4 Ve s s e l Pl a t e ( S_A-53 3 Gr. B C1. 1) and Formina (SA-50s C1. 2) Weld RA2:

RT assumed name as for base material. Wald procedure qualification test NUT requirements ladicate this assnaption is valid.

2.5 Boltion Material (SA-544 Gr. B24);

CVN 43 f t-1b and 25 WLE (Nils L.teral Expansion) are required at no higher than preload temperature or Lowest Service Temperature (LST)

Usual data available - CVN f t-lb ana MLE at +10*F LST prediction method -

If preceding CVN requirements are set at test temperature, then it is LST.

If at least 30 f t-lb, but less than 45 (t-1b and 25 MLE, ase met at test

, temperature, then add 60*F to the test t emperature f or LST.

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.l FS 28 '840 2 5 9 7 75 Attachment III HCGS FSAR 10/83 OUESTION 22A 16 (SECTION 3.8.4) f Indicate whether satorials, fabrication, welding, and quality l control of the spent fuel racks are in conformance with i applicable provisions of subsection NF of the ASME code. If not, identify and justify the deviations.

j 1 RESPONSE ,

7 Ths spqnt fuel rack specification requires conformanes yf M [

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Subsectaan arreautrements for Class 3 c-==e'fiipports. Tne <

purchase order is scWto be tsartred ;,n December 1993. This

, response will be verfiri b ... ..vtactens, i' = h will be provided after p a or the purchase order. Table .i-iii.si,,betn snow Subsection NF as the principal code for the

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sensary as&a sectiences ser specific ser seecripties of there

_Iss1 Lee. _ asseelsese_Etaterie_____ agggegesses__ _ _ _ . alecease4 . -

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Category I stracteres shall Category I stracteres are de-be designet la a=cereence signed te seeereesse eith  !

sith specifisettee tcI 349 specificaties ACI 386-79.

es segeested by Segolatory l Seide 1.142.

l II.2 8. 8.1 Confermance to Segolatory h 8erseems, le seat.

, Seteos 1.18, 1.55 med 1.94. sith gegelatory esides 1.14, , j l.55, and 9.94. '

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Deelge resorte are acceptable sofficient laternaties is Ef it centales the tafermetten available la forme other

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Deelge report ie coasteered soffaciest information is CD j e=ceptable if it satisfies the ave 11ehte le forse other getes1Esos of aposedia C to thee these est11eed le j -

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seapeeeste aboots he 1 edged Apossets & ar set 3.9.3. oi by eespeciese eith eseittees j stated le Aossedis & et set 3.9.3.

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' HCGS FSAR 1/84 g W.} l facility. They are not shear walls and are designed to the working stress method of UBC, as listed in Table 3.8-7.

3.8.4.6 Materials, Quality Control, and Special Construction Techniaues

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Materials, quality control, and special construction techniques are discussed in Section 3.8.6.

J 3.8.4.7 Testino and Inservice Inspection Reauirements

! Testing and inservice inspection are not required for Seismic

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' SRP 3.8.4.8 ^ Rule Review x

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{ 3. 8. 4. 8.1 - Concrete Design i

Acceptance Criteria II.2 of SRP 3.8.3 and 3.8.4 requires that 4 Category I structures be designed in accordance with i Specification ACI 349 as augmented by Regulatory Guide 1.142. ,

( , The HCGS design was based on the requirements of Specification '

ACI 318-71. /~ N i .

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The Category I structures concrete design for HCGS began prior to -

the issue of Specification ACI 349 (1976). As a result, all ,

, concrete design is based on using Specification ACI 318-71 with l i the following clarifications:  !

A review of the design of the MCGS Seismic Category I structures indicates that there is no impact due to differences in the structural acceptance criteria between ACI 318-71 and ACI 349-76 as augmented by Regulatory Guide 1.142.  ;

1 The load combinations used are in conformance with the following SRP sections except that the 0.9 load factor on dead load as j required by ACI 349-76 was not used:

Structures SRP Section l l Primary Containment 3.8.3.II.3.b.

Internal Concrete Structures Other Seismic Category I 3.8.4.II.3.b.

Concrete Structures Based on parametric analyses, an adectuate design margin exists to compensate for the effects of the re(uced dead load factor. ,,.

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• Table 3.8-18 provides a comparison of the allowable ductility l

! ratios used for design of the concrete structural components l subjected to impactive and impulsive loadings and the criteria i

! outlined in Appendiz C of ACI 349 as modified by Regulatory l l

Guide 1.14.2.

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II Except for flexural beams and slabs subjected to impactive loads,

!  ! the allowable ductility ratios used in the design are less than i or equal to those in the Regulatory Guide. The allowable -

i ductility ratios for beams and slabs used in design are based on l the evaluation of test data reported in References 3.8-5 and l l 3.8-6 and tests performed by the Architect-Engineer.

! I l lI The test results consistently demonstrate that actual ductility  !

l! ratios in excess of 50 are reached prior to failure. Therefore, i : by limiting the values to 10 for beams and 30 for slabs, the <

j[ design is conservative. Furthermore, the flesural members are  :

i designed to meet additional reinforcing requirements (See Table l

3.8-18) to ensure ductile behavior.  ;

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j 3.8.4.8.2 Structural Steel Design i Table 3.8-19 provides a comparison of the allowable ductility ratios used for design of structural steel subjected to impactive j

l and impulsive loading, and the criteria outlined in Appendia A of '

NUREG-0800, SRP Section 3.5.3. Except for flesure in seems  :

subjected to impactive loads (other than the tornado missiles) '

, and asial tension members subject to impulsive loads, the 1

!, ductility ration are essentially identical. Based on the  !

j recommendations provided in Refierences 3.8-5 and 3.8-4 and tests j performed by the Architect-Engineer, it has been demonstrated i

that steel sesbers under f1tuural loads can sustain higher

ductility ratios (on the order of 30) without collapse.  !

j Therefore, a limiting value of 20 used in the design is l conservative. Furthermore, additional design and fabrication features (such as bon sections, lateral bracings, NDE, etc.) are ,

incorporated in the flesural members to preclude buckling and to  !

4 ensure material quality. ,

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to compression loads, the BCGS limit of 3 is always conservative  !

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Attcchm:nt IV HCGS FSAR 1/84 1 f, QUESTION 430.94 (SECTION 9.5.4 ) i In Section 9.5.4.2.6 of the PSAR you state that the emergency flood protected truck till connection for the fuel oil storage -

tanks is located inside the auxiliary building at Elevation 102 feet (plant grade level). In Section 3.4, Table 3. 4-1 you state that the design flood elevation for the D/G building is 120.4 feet with the still water height at 113.8 feet. Provide the followings

a. Describe or provide adequate drawings to show the location of the emergency fuel oil storage tank fill connection.

b. Assuming the emergency fill connection must be used to refill the f uel oil storage tanks. Describe how f uel oil will be delivered to the site during flood conditions and describe the procedures that will be used in refilling the storage tanks during flood conditions and non-flood conditions. The procedures should include fuel hose routing and f tre watcher.

c. Describe how flood water is prevented from entering the building during refueling operations. (SRP 9.5.4, Parts I, II & III).

RESPONSE

Th6 diesel fuel oil emergency fill line is located in the auxiliary building at floor elevation 102 feet-0 inches and a center line elevation of 106 feet-6 inches. The emergency diesel fuel oil fill connection is located in an area which is flood protected by the auxiliary building main service entry doors.

The location of the diesel fuel oil connection is shown on Figure 430.94-1, reference Figure 1.2-35 for location relative to watertight door.

Response to Item (b) will be provided in April, 1984.

( Leakage through the door seals is removed by drainage systems in the bu ilding . The flood doors are capable of withstanding the flood height as described in Section 3.4 and Table 3.4-1.

The diesel fuel oil tanks are designed for a seven day supply with the diesel generators operating at full capacity, reference Section 9.5.4.3. It is not anticipated that localized flooding will prevent at_ somattime during thid#y period. fy a t re v.%f ue li ng fThecedure opur(fig rom occuri ngS.4.// as'/rea /4e. f a rif /,,,j e /

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Attachment V HCGS FSAR 10N#ilSD0 l

OUESTION 640.9 (SECTION 14.2.12) l Modify FSAR Subsect19n 14.2.12.1.29 (KC-Fire Protection - Deluge) to provide assurance that:

1. Upon automatic sprinkler actuation, adequate drainage in t;he affected spaces is provided to preclude flooding (including expected hand-held hose volume).

2. A walk-down of plant equipment is conducted to identify potential incidences where the actuation of ftre suppression systems could cause damage to or inoperability of systems important to safety. ,

See IE Information Notice 83-41: Actuation of Fire Suppression System Causing Inoperability of Safety-Related Equipment, -

June 22, 1983. *

RESPONSE

~ . . , _c FThe results of our findings on adequacy of the drainage to preclude flooding, upon automatic sprinkler actuation, will be ,

available March 19E4. _

/ Section 14.2.12.1.29.b has been revised to include a prerequisite walkdown of the ftre protection system to identify potential areas where the fire protection system could cause damage.

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2. The system responds to simulated fire signals.

3. The refrigeration system operates to maintain pressure and temperature as specified by the manufacturer's technical instruction manual.

14.2.12.1.29 KC-Fire Protection - Deluge

a. Objective The test objective is to verify the capability of the fire protection system to deliver water to the sprinkler system, pre-action and deluge systems, hose .

stations, and hydrants at rated pressure and flow.

b. Prerequisites

1. Component tests have been completed and approved.

2. System instrumentation has been calibrated and

approved. ),

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3. AC and de power are available.

4. The diesel fire pump local fuel oil storage tank is in service.

5. Adequate fire protection water supply is available.

6. A walkdown has been performed to identify components or areas that may be susceptible to 7, ,,gg 7 damage due to actuation of the deluge system.

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1. 8## c. Test Method

1. All valves, controls, alarms, interlocks, and '

logic are checked for proper operation.

2. Normal system flow paths are verified.

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