ML20083C516

Forwards Preliminary FSAR Change Re Fuel Sys Design to Ref Gessar Ii.Change Will Be Incorporated Into Amend 4 to FSAR Scheduled for Jan 1984

ML20083C516
Person / Time
Site:	Hope Creek
Issue date:	12/14/1983
From:	Mittl R Public Service Enterprise Group
To:	Schwencer A Office of Nuclear Reactor Regulation
References
NUDOCS 8312230171
Download: ML20083C516 (8)

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O PSEG Public Service Electric and Gas Company 80 Park Plaza, T16D Newark, N.J. 07101 201/430-8217 Robert L. Mitti General Manager Licensing and Environrnent December 14, 1983 Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20014 Attention: Mr. Albert Schwencer, Chief Licensing Branch 2 Division of Licensing Gentlemen:

HOPE CREEK GENERATING STATION DOCKET NO. 50-354 FSAR REFERENCE REVISION FOR FUEL SYSTEM DESIGN Public Service Electric and Gas is currently revising Section 4.2, " Fuel System Design," of the HCGS FSAR, to reference the General Electric Standard Safety Analysis Report (GESSAR II). The fuel system design for Hope Creek is identical to that already reviewed by the staff for GESSAR II. Methods and criteria used to evaluate fuel system performance are also identical to those used for GESSAR II. The Core Performance Branch has stated that the reference to GESSAR II permits them to incorporate the results of the GESSAR review in accordance with Standard Review Plan 4.2 (NUREG-0800) into NUREG-0979, thus signif-icantly reducing NRC staff efforts which would be required to review HCGS FSAR Section 4.2. The preliminary FSAR change is attached for your convenience. This change will be formally transmitted to the staf f as part of Amendment 4 of the HCGS FSAR presently scheduled for submittal in January 1984.

Very truly yours,

. s' ,l gl '

1 831E230171 831214 PDR ADOCK 05000354 A pyg '

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Attachment

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The Energy People 1

Director of Nuclear 12/14/83 Reactor Regulation CC: D. H. Wagner USNRC. Licensing Project Manager W. - H . Bateman

- USNRC Senior Resident Inspector.

C. H. Berlinger US NRC Core Performance Branch R. O. Meyer

.US NRC Core Performance Branch

.BR 10 01/02-B

B

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HCGS'FSAR TABLE 1.6-1 (cont) Page 3 of 10 Report Referenced in Number Title FSAR Section NEDE-21156 Supplemental Informati for Plant 4.4 Hodification to Elimit c ' agnificant In-Core Vibration, January 1976.

NEDE-21354'-P BWR Fuel Channel Mechanical Design 3.9 and Deflection, September 1976.

NEDE-23014 HEX 01 User's Manual, July 1976. 15.2 NEDE-23786-P Fuel Rod Prepressurization, March 1978. 4.2' NEDE-24222 Assessment of BWR Mitigation of ATWS 15.8 f (NUREG-0460 Alternate No. 3),

Volume 1, May 1979; Volume 2, December 197.9.

NEDE-24226-P Evaluation of Control Bl'ade Life with 4.2 Potential Loss of B.C, December 1979.

($ NEDE-24834 Hanford 2 Crimped CRD Hydraulic 36 1(

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Withdrawal Lines, (Proprie,tary).

NEDO- 10029 An Analytical Study on Bri~ttle 5.3 Fracture of GE-BWR Vessel Su'bjected to the Design Basis Accident, July 1969.

NEDO-10173 Current State of Knowledge, High 11.1

-' Performance BWR Zircaloy-clad UO, Fuel, May 1970.

'NEDO-10299-A Core Flow Distribution in a Modern 4.4 Boiling Water. Reactor as Measured in Monticello, October 1976.

NEDO-10320 The Generai_ Electric Pressure 6.2 Suppression Containment Analytical Model, April 1971; Supplement 1, -

May 1971.

NEDO-10349 Analysis of Anticipated Transients 15.8 Without Scram, Marc,h 1971.

WE 6c B o t \ -P' A &*M EO ' YN APf lh fo r Oc Fvst H.H

( U*d QvcA rev.%'on i AIEb E- nolt - P- A-US &nd. EluJr:e LLL Appl.& 4. 4 for Realce iv e-l tf Id

• [.900) afpoveel, Un;M ttsNs.'on ) 2 kk h

4 HCGS FSAR

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k TABLE 1.6-1 (cont) Page 2 of 10 Report Referenced in Number Title FSAR Section GEAP-5620 Failure Behavior in ASTM A1063' Pipes 5.2 Containing Axial Through-Wall Flows, April 1968 GEAP-10546 Theory Report for Creep-Plast Computer 4.1  !

Program, January 1972.

KAPL-2170 Hydrodynamic Stability of a Boiling 4.4 Channel, October 1961. ,

KAPL-2208 Hyd'odynamic r Stability of a Boiling 4.4 Channel, Part 2, April.1962.

KAPL-2290 Hydrodynamic Stability of a Boiling 4.4

. Channel, Part 3, June 1963.

' KAPL-3070 Hydrodynamic Stability of a Boiling 4.4 Channel, Part 4, August 1964.

[ _ . ; KAPL-3072 Reactivity Stability of a Boiling Water Reactor, Part 1, September 1964.

4.4

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KAPL-3093 Reactivity Stability of a Eoiling 4.4 Water Reactor, Part 2, March 1965.

NEDE-10313 PDI - Pipe Dynamic Analysis Program 3.6 for Pipe Rupture Movement (Proprietary

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Filing).

NEDE-10958 General Electric BWR Thermal Analysis 15.0 Basis (GETAB): Data, Correlation and Design Application, November 1973.

NEDE-20566-P-A Analytical Model for Loss-of-Coolant 4.2 Analysis in Accordance with 10 CFR 50, Appendix K, November 1975.

,, NEDE-20944-P BWR/4 and BWR/5 Fuel Design, Proprietary Versions, October 1976.

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NEDE-20944-1P. BWR/4 and BWR/5 Fuel Design, A48, 4 '. 3 Amendment 1, (only BWR/4&5,) Apr January 1977.

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.' IS HCGS FSAR //G3 CHAPTER 4 REACTOR TABLE OF CONTENTS 1

Section .

. Title 4.1

SUMMARY

DESCRIPTION 4.1.1 Reactor Vessel 4.1.2 Reactor Internal Components 4.1.2.1 Reactor Core .

4.1.2.2 Shrou'd 4.1.2.3 Shroud Head and Steam Separator Assembly 4.1.2.4 Steam Dryer Assembly 4.1.3 Reactivity Control Systems 4.1.3.1 Operation 4.1.3.2 Description of Control Rods 4.1.3.3 Supplementary Reactivity Control l 4.1.4 Analysis Techniques l 4.1.4.1 Reactor Internal Components 4.1.4.2 Fuel Rod Thermal Design Analyses 5 4.1.4.3 Reactor Systems Dynamics i 4.1.4.4' Nuclear Engineering Analysis .

! 4.1.4.5 Neutron Fluence Calculations i 4.1.4.6 Thermal-Hydraulic Calculations 4.1.5 References 4.2 FUEL SYSTEM DESIGN

, [4. 2R Design Bases tion and Design Drawin,.

l 4.2.2 l

'4.2.2.1 Reactivi -1 assembly (Control Rods) -

4.2.2.2 y Contro embly Evaluation i 4.2.3 Design Evaluation l .A . Testing, Inspection and Surveilla 4.2. E>1 References

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4.3 NUCLEAR DESIGN 4.3.1 Design Bases 4.3.2 Description 4.3.2.1 Nuclear Design Description 4.3.2.2 Power Distributions 4.3.2.3 Reactivity Coefficient l 4.3.2.4 Control Requirements 4.3.2.5 Control Rod Patterns -and Reactivity Worths 4.3.2.6 Criticality of Reactor During Refueling 4.3.2.7 Stability 4.3.2.8 Vessel Irradiations 4.3.3 Analytical Methods

. IL Amendment /f 4-i

• 10 HCGS FSAR K/83 1

4.2 FUEL SYSTEM DESIGN (

iitJWRT '

m TTAS4ED he fo'rmat of this section corresponds to Standard Review lan 4.2 in NUREG-0800. Most of the information is presented by

' r ference to GESTAR II (Reference 4.2-1).

4.2. DESIGN BASIS l Referene s to design bases are given in Subsection A.4. .1 of i Reference 4.2-1.

1 4.2.2 DE RIPTION AND DESIGN DRAWINGS l References to th fuel system description and sign drawings are given in Subsecti A.4.2.2 of Reference 4.2 .

ntrol'Assembl'y (Co

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4.2.2.1 Reactivity rol Rods) l The control rod descripti is given ' Subsection 2.2.4 and is shown on Figures 2.6a, 2.6b, and 2.7 of NEDE-20944-P-1 (Reference 4.2-2).

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4.2.2.2 Reactivity Control s bly Evaluation l

The control rod evaluatio is given Subsection 2.3.3 of

! Reference 4.2-2.

I 4.2.b DESIGN EV UATION l Compliance wit the design bases is discussed i Subsection A .2.3 of Reference 4.2-1.

4.2.4 TESTING, INSPECTION AND SU,RVEILLANCE PLANS '

l Des iptions of fuel assembly testing, inspection and s veillance are referenced in Subsection A.4.2.4 of ference 4.2-1.

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4.2-1 Amendment A

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HCGS FSAR 3/83

. 5 REFERENCES

) l 4.2-1 " General Electric Standard Applic

• for Reactor F ' " including the " Unite es Supplement,"

NEDE- 1-P-A and NE 011-P-A-US (latest approved r i .

4.2-2 4 and BWR/5 Fuel De "

NEDE-20944-P-1 (Proprietary).and NEDO-20944 , ober 1976, and

" Amendment 1," January 1977.

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INSERT The fuel system design for the HCGS is identical to that which the NRC reviewed and approved for GESSAR II (Reference 4.2-1). Methods and criteria used to evaluate fuel system performance are also identical to those used for GESSAR II. The results of the NRC review of Section 4.2 of GESSAR II documented in References 4.2-2 and 4.2-3 are therefore applicable to the HCGS.

4.

2.1 REFERENCES

4.2-1 General Electric Standard Safety Analysis Report.

. . Docket No. 50-447

' 4.2-2

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NUREG-0979, " Safety Evaluati.on Report Related to the Final Design Approval of the GESSAR II BWR/6 Nuclear Island Design", April, 1983

~ 4.2-3 NUREG-0979 (Supplement No.1), " Safety Evaluation Report Related to the Final Design Approval of the GESSAR II BWR/6 Nuclear Island Design'.', July,1983 O

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