ML18100A756
| ML18100A756 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 12/07/1993 |
| From: | Hagan J Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NLR-N93193, NUDOCS 9312140172 | |
| Download: ML18100A756 (8) | |
Text
... ***
~..*
Public Service Electric and Gas Company Joseph J. Hagan Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1200 Vice President - Nuclear Operations DEC 0 7 1993 NLR-N93193 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
CYCLE 12 RELOAD ANALYSIS FACILITY OPERATING LICENSE DPR-70 UNIT NO. 1 SALEM GENERATING STATION DOCKET NO. 50-272 Salem Unit 1 has completed its eleventh cycle of operation on October 2, 1993.
The burnup at the end of Cycle 11 was 13035 MWD/MTU.
The startup of Cycle 12 is scheduled for December 4, 1993.
The intent of this letter is to inform you of PSE&G's plans regarding the Salem Unit No. 1 cycle 12 reload core which is expected to achieve a burnup of 15100 MWD/MTU.
The Cycle 12 reload will utilize two regions of fresh fuel (Figure 1).
The first region consists of 4 Region 14A assemblies enriched to 4.00 w/o u235 with no Integral Fuel Burnable Absorber (IFBA) rods and 28 Reg1on 14A assemblies enriched to 4.00 w/o u235 containing 64 1.5X IFBA rods.
The second region consists of 20 Region 14B assemblies enriched to 4.40 w/o u235 containing 64 1.5X I~BA rods.
The loading contains a total or 512 fresh burnable absorber rodlets, 3072 1.5X IFBA rods and 768 Zircaloy-4 rodlets arranged as shown in Figure 2.
The design of the Region 14 fuel assemblies is the same as the Region 13 assemblies except for the use of; (1) rotated mixing-vane grids, (2) keyless/cuspless top nozzle assembly, (3) a composition bottom nozzle, (4) an extended burnup bottom grid, (5) a protective coated cladding, and (6) the use of 1.5X IFBA.
Two Region 10 fuel assemblies, K24 in core location M-08 and K41R in core location L-06, have been reconstituted by replacing a damaged fuel rod with a stainless steel filler rod in pin locations K-10 and N-16, respectively.
In addition to the stainless steel filler rod, K41R has also been recaged with a VANTAGE 5H skeleton due to previous grid damage to Grid #1.
Stainless steel filler rods have been used as replacements for fuel rods in previous Salem Cores and do not adversely affect the
- 13on /1 J:*'.~-
r-- - __
v_L;r __ _
9312140172 *9a1207- -*-n*--...
PPDR ADOCK 050Q02'72
- }
- -PDR LJ
Document Control Desk NLR-N93193 2
DEC 0 7 1993 mechanical integrity of the assembly.
The nuclear evaluation for fuel rod reconstitution has been performed in accordance with approved NRC codes and methods (References 3 and 4).
The evaluation has assessed the safety significance of fuel rod reconstitution and recaging and has assured that a core containing assemblies with reconstituted stainless steel filler rods and recaged assemblies meet the design criteria for existing fuel design and that the core loading pattern will have no adverse impact on parameters used in the accident analysis for Cycle 12.
Due to recent observations of fuel rod clad fretting wear failures in V5H assemblies during the Salem Unit 2 7th refueling outage, 24 Zircaloy-4 damper rods will be inserted into 32 burned V5H assemblies loaded adjacent to the core baffle (Figure 2).
The purpose of the Zircaloy-4 damper rods is to minimize flow-induced vibration thereby reducing the possibility of fuel rod clad fretting failures.
For V5H feed fuel assemblies (Region 14), alternate grids are rotated 90 degrees clockwise in order to minimize the susceptibility of flow-induced fuel assembly vibration.
There are no other physical changes to the grids or their axial positions.
Westinghouse has completed the safety evaluation for the Cycle 12 reload core design utilizing the methodology described in Reference 1.
Based on this methodology, those incidents analyzed and reported in the Salem UFSAR (Reference 2) that could potentially be affected by the fuel reload are addressed.
The dropped RCCA incidents were conservatively evaluated assuming the removal of the Negative Flux Rate Trip (NFRT) function based on the approved dropped rod methodology (Reference 5).
- However, the evaluation for the dropped RCCA incidents with Negative Flux Rate Trip (NFRT) function, based on approved methodology, are also bounded.
The potential control rod system failure described in LCR 93-21 (Reference 6) was evaluated in the safety evaluation for the Salem Unit 1 Cycle 12 reload.
The analysis performed specifically for Salem Unit 1 Cycle 12 asymmetric rod withdrawals at power meets the Condition II acceptance criteria.
The boron concentration reduction in the Boric Acid Tanks (BATs) identified by LCR 93-17 has been addressed for Salem Unit 1 cycle 12 (References 7 and 8).
The reduced concentration eliminates the need to heat trace the boric acid makeup system piping and equipment to prevent boric acid precipitation.
The design basis for the BAT volume and concentration requirement is to provide sufficient boric acid to borate the Reactor Coolant System to cold shutdown, assuming the control assembly with the highest
Document Control Desk NLR-N93193 3
DEC 07 il5ru worth is stuck fully withdrawn.
This design basis is not modified by reducing the BAT concentration.
Boration will be performed during the cooldown process, instead of prior to cooldown.
The required method changes which allow boration during cooldown were provided as part of LCR 93-17.
The core physics assumptions used in the license change analysis remain bounding for Salem Unit 1 Cycle 12.
Large Break LOCA analyses have been traditionally performed using a symmetric, chopped cosine axial power shape.
Recent calculations have shown that there was a potential for top-skewed power distributions to result in peak cladding temperatures (PCT) greater than those calculated with a chopped cosine axial power distribution.
Westinghouse has developed a process, which was applied to the reload for Salem Unit 1 Cycle 12, that ensures that the cosine remains the limiting power distribution, by defining appropriate power distribution surveillance data.
This process, called the Power Shape Sensitivity Model (PSSM), is described in topical report WCAP-12909-P and further clarified in ET-NRC-91-3633.
The safety evaluation states that all Cycle 12 kinetics parameters, control rod worths, and core peaking factors meet the current limits with the exception of the normalized trip reactivity insertion rate which is slightly different from the current limit.
The normalized trip reactivity insertion rate was compared to the previous analyses and evaluated for those accidents affected.
The conclusions of analyses in the Salem UFSAR (reference 2) were demonstrated to remain applicable.
A review of the Salem Unit 1, Cycle 12 Reload Safety Evaluation (RSE) has been performed relative to the impact of this RSE on the Salem Unit 1 Technical Specifications (Reference 10).
As a result of this review, no Technical Specification changes are required based on the subject RSE for Cycle 12 operations.
The Radial Peaking Factor Limit Report previously submitted in Reference 9 for Salem Unit No. 1 Cycle 12 is still applicable for this revised pattern.
PSE&G has reviewed the basis of the Cycle 12 reload analysis and the Westinghouse Reload Safety Evaluation Report with Westinghouse.
We have determined that all the postulated events are within allowable limits and that no unreviewed safety questions as defined by 10CFR50.59 are involved with this reload.
Document Control Desk NLR-N93193 4
DEC 07 1f993 The reload core design will be verified during the startup physics testing program.
The program will include, but is not limited to the following tests:
- 1.
Control rod drive tests and drop time measurements.
- 2.
Critical boron concentration measurements.
- 3.
Control rod bank worth measurements.
- 4.
Moderator temperature coefficient measurements.
- 5.
Power distribution measurements using the incore flux mapping system.
Should you have any questions regarding this transmittal, please do not hesitate to contact us.
Sincerely, Attachments
Document Control Desk NLR-N93193 5
C Mr. T. T. Martin, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. J. c. Stone, Licensing Project Manager U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. c. Marschall (S09)
USNRC Senior Resident Inspector Mr. K. Tosch, Manager, IV NJ Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 DEC 0'7 l993
References:
- 1.
Davidson, S. L. (Ed.), et. al., "Westinghouse Reload Safety Evaluation Methodology," WCAP-9273-NP-A, July 1985.
- 2.
Salem Units 1 and 2 Updated Final Safety Analysis Report, USNRC Docket Numbers 50-272 and 50-311, July 22, 1992.
- 3.
Slagle, w. H. (Ed.), "Westinghouse Fuel Assembly Reconstitution Evaluation Methodology", WCAP-13060-P, September 1991.
- 4.
Letter from A. c. Thadani (NRC) to s. R. Tritch (Westinghouse), Acceptance for Referencing of Topical Report WCAP-13060-P, "Westinghouse Fuel Assembly Reconstitution Evaluation Methodology," March 30, 1993.
- 5.
Haessler, R. L., et. al., "Methodology for the Analysis of the Dropped Rod Event," WCAP-11394-A, January 1990.
- 6.
LCR 93-21, NLR-N93098, Request for Emergency License Amendment Rod Control System, Salem Generating Station Units 1 and 2, Facility Operating License Nos. DPR-70 and DPR-75, Docket Nos. 50-272 and 50-311.
- 7.
LCR 93-17, NLR-N93077, Request for Amendment, Boron Concentration Reduction, Salem Generating Station Unit Nos.
1 and 2, Facility Operating Licenses Nos. DPR-70 and DPR-75, Docket Nos. 50-272 and 50-311, June 11, 1993.
- 8.
NRC Amendment 145, "Boron Concentration in Boric Acid Tanks Reduced, Salem Nuclear Generating Station, Unit No. 1, (TAC NO. M86723)," October 15, 1993.
- 9.
Letter from J. Hagan (PSE&G) to United states Nuclear Regulatory Commission, "Cycle 12 Radial Peaking Factor Limit Report, Salem Generating Station, Unit No. 1, Docket No.
50-272, 11 October 26, 1993.
- 10.
Salem Unit 1 Technical Specifications, USNRC Docket Number 50-272.
0 SALEM UNIT 1 CYCLE 12 REDESIGN R
p 9
l3A 12 148 138 13A 138 148 14A 12 138 148 138 13A 12 148 13.l\\
9 LEGEND R
r Rev. I 0129/93 5: I Opm N
M L
I I I
12 9
13A 148 138 l4A 12 l4A 138 l3A 12 13A llA 14A 12 10 12 14A llA 138 10 14A 12 l4A llA 14A 12 13A 12 13A llA 14A 138 13A 138 14A 12 9
l3A 148 12 Figure 1 Core Loading Pattern Salem Unit 1 - Cycle 12 K
J H
G F
I I
I 138 138 UA 138 138 l3A 148 12 148 l3A l4A 12 138 12 14A 12 l4A 10 14A 12 13A llA 14A llA 13A 12 13A 12 13A 12 13A 12 148 12 13A 12 148 10 148 12 13A 12 148 12 13A 12 1*3A 12 10 12 13A llA 14A llA 13A 12 14A 10 14A 12 14A 12 138 12 14A 13A 148 12 148 l3A 138 138 14A 138 138 REGION IDENTIFIER llECONS'tTfUTED ASSEMBLY 24 NOVEMBER 1993 E
D c
8 A
I 12 1
148 l3A 9
2 12 l4A 138 9
l3A 138 14A l3A 4
llA 13A 12 148 12 5
13A 12 14A 13A 138 6
llA 14A 12 148 138 7
14A 10 138 12 14A 8
llA 14A 12 148 138 9
13A 12 14A 13A 138 10 llA. 13A 12 148 12 11 13A 138 14A 13A 12 12 14A 138 9
148 l3A 10 14 12 15
0 SALEM UNIT 1 CYCLE 12 REDESIGN NOVEMBER 1993 @
R p
24WD 4P 24WD 64I 24WD 12P 24WD 64I 12P 24WD 64I 24WD 4P 24WD 64I 24WD N
12P 64I 12P 64I 12P 64I 12P 64I M
Figure 2 Burnable Absorber and Source Rod Locations Salem Unit 1 - Cycle 12 L
K J
H G
F E
D I
I I
I I
I I
24WD 24WD 24WD 24WD 24WD 24WD 4P 12P 12P 4P 24WD 24WD 64I 64I 64I 64I 12P 12P 12P 12P 4SS.A 64I 64I 64I 64I 12P 12P 64I 64I 12P 64I 12P 12P 12P 64I 64I 64I 12P 12P 12P 12P 64I 64I 64I 64I 12P 12P 12P 64I 64I 64I 12P 64I 12P 12P 64I 64I l2P l2P l2P l2P 4SSA 64I 64I 64I 64I 4P l2P l2P 4P 24WD 64I 64I 64I 64I 24WD 24WD 24WD 24WD 24WD 24WD 24WD go TYP£ TOTAL c
12P 64I 12P 64I 12P 64I 12P 64I NP ___.(NUMBER OF PYREX RODLETS).____
512 NL __.(NUMBER OF IFBA RODS)*----------------* 3072
- SSA._.(NUMBER OF SECONDARY SOURCE RODLm)_
8 NWO._(NUMBER OF DAMPER ROD RODLE'J'S)___
761 Rev. I 0/29193 S: I Opm 25 B
A 1
2 3
24WD 4
4P 64I 24WD 5
24WD 6
12P 24WD 7
64I 8
12P 24WD 9
64I 24WD 10 4P 24WD 11 64I 24WD 12 13 15