ML18101A735

Submits Summary of Plant Startup & Power Escalation Testing for Ninth Cycle of Operation

ML18101A735
Person / Time
Site:	Salem
Issue date:	05/15/1995
From:	Labruna S Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LR-N95077, NUDOCS 9505260180
Download: ML18101A735 (8)

Text

• Public Service Electric and Gas Company Stanley LaBruna Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1700 Vice President - Nuclear Engineering MAY 151995 LR-N95077 United States Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Gentlemen:

CYCLE 9 RELOAD ANALYSIS AND STARTUP TESTING FACILITY OPERATING LICENSE DPR-75 UNIT NO. 2 SALEM GENERATING STATION DOCKET NO. 50-311 In accordance with Salem Generating Station Unit 2 Technical Specification 6.9.1.1, Public Service Electric and Gas Company (PSE&G) submits the summary of plant startup and power escalation

• testing for the ninth cycle of operation.

Salem Unit 2 completed its eighth cycle of operation on October 13, 1994. The burnup at the end of cycle 8 was 14,050 MWD/MTU.

The startup activities for Cycle 9 were completed on March 22, 1995. The Salem Unit 2 Cycle 9 reload core is expected to achieve a burnup of 15,300 MWD/MTU.

The Cycle 9 reload will utilize two regions of fresh fuel. The first region consists of 52 Region llA fuel assemblies (4.0 w/o U235, 28 assemblies containing 64 Integral Fuel Burnable Absorber (IFBA) pins, 16 assemblies containing 80 IFBA pins, 8 assemblies containing 104 IFBA pins). The second region consists of 12 Region llB fuel assemblies (4.4 w/o U235, 8 assemblies containing 104 IFBA pins, 4 assemblies containing 48 IFBA pins). Also 28 Region 12 fuel assemblies built for Unit 1 Cycle 10 which were used during Unit 2 Cycles 7 and 8 will be reinserted for a third cycle of irradiation. A total of 21 assemblies were reinserted from the spent*fuel pool (Cycle 7 irradiation; 17 thrice burned Batch 7A standard assemblies and 4 twice burned Batch SA V5H assemblies). The remaining assemblies were retained from Cycle 8 (16 twice burned Region 9A (3.8 w/o U235), 8 twice burned Region 9B (4.0 w/o U235), 24 once burned Region lOA (4.0 w/o U235), 32 once burned Region lOB (4.4 w/o U235)). The Cycle 9 core contains 448 fresh burnable absorber rodlets and 32 Zircaloy-4 vibration dampener assemblies (24 rodlets per assembly). These Zircaloy-4 vibration dampener assemblies are inserted in burned V5H (with no alternate grid rotations) assemblies located in the periphery of the core in order to minimize the possibility of 9505260180 950515 PDR ADOCK 05000311 P PDR

Document control Desk 2

• MAY 1 51995 LR-N95077 flow induced vibration that could cause fuel rod clad fretting.

In addition, in order to utilize all the available Zircaloy-4 vibration dampener assemblies, four inboard V5H assemblies will also contain vibration dampener assemblies.

Region 11 of Cycle 9 introduces the use of ZIRLO' fuel cladding and ZIRLO' mid-grid assemblies. ZIRLO' fuel rods grow approximately half as much as those of standard Zircaloy-4, with irradiation. ZIRLO' also provides increased corrosion resistance over that of standard Zircaloy-4.

The mechanical design of the Region 11 fuel assemblies is the same as Region 10 assemblies except that the Region 11 assemblies incorporate the following fuel design improvements: (1) keyless/cuspless top nozzle with modified spring tang, (2) shorter guide thimble/instrumentation tube fabricated with ZIRLO' and shorter fuel rods clad with ZIRLO', (3) mixing vane grids (mid-grids) fabricated with ZIRLO', (4) rotated mixing vane grids, (5) variable pitch plenum spring, (6) extended burnup bottom grid design, (7) protective coated cladding with double grit blasting of the fuel rod, (8) reduced fuel rod to bottom nozzle gap, (9) IFBA coated fuel pellets with increased B-10 loading, and (10) axially aligned, reduced stack length, Pyrex Burnable Absorbers. Changes (1) and (10) have been introduced so that parts and manufacturing processes may be standardized and the product will be more uniform and consistent. Change (4) has been incorporated to eliminate the fuel assembly vibration occurring in assemblies located face adjacent to the core baffle.

Changes (2), (3) and (5) through (9) are fuel upgrade changes to enhance fuel performance.

Westinghouse has completed the safety evaluation of Cycle 9 reload core design utilizing the methodology described in Reference 1. Based on this methodology, those incidents analyzed and reported in the Salem Updated Final Safety Analysis Report (UFSAR, Reference 2) that could potentially be affected by the fuel reload are addressed. The dropped Rod Cluster Control Assembly (RCCA) incidents were conservatively evaluated assuming the removal of the Negative Flux Rate Trip function based on the approved dropped rod methodology (Reference 3). However, the evaluation for the dropped RCCA incidents with the Negative Flux-Rate Trip function, based on approved methodology, is also bounded.

Large Break Loss of Coolant Accident 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 95-4933

Document Control Desk 3 MAY 1 51995 LR-N95077 Unit 2 Cycle 9 that ensures 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 a topical report (WCAP-12909-P) and further clarified in ET-NRC-91-3633, both of which are currently under NRC review.

The safety evaluation states that all Cycle 9 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. Slow transients are relatively insensitive to the trip reactivity insertion rate and the conclusions of the UFSAR remain valid for these transients. The fast transients such as loss of forced reactor coolant flow and locked rotor which can be sensitive to the normalized trip reactivity insertion rate were reanalyzed using more conservative values.

than the current limit reactivity insertion rates. The trip reactivity curve used for these analyses does bound the Cycle 9 calculated reactivity insertion rate values.

For Cycle 9 the boron concentration in the Boric Acid Tanks (BATs) has been reduced from 12.5% to 4% in order to eliminate heat tracing of the boric acid makeup system piping and equipment needed to prevent boric acid precipitation (Reference 4). At the reduced concentration, the ambient temperature in the station auxiliary building is sufficient to prevent precipitation. The core physics assumptions used for the license change analysis to determine the system boration requirements during cooldown remain bounding for Salem Unit 2 Cycle 9.

A review of the Salem Unit 2, Cycle 9 Reload Safety Evaluation (RSE) has been performed relative to the impact of this RSE on the Salem Unit 2 Technical Specifications (Reference 5). As a result of this review, no Technical Specification changes are required based on the RSE for Cycle 9 operation.

The Radial Peaking Factor Limit Report for Salem Unit 2 Cycle 9 was submitted in Reference 6.

PSE&G has reviewed the basis of the Cycle 9 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.

Therefore, based on this review, application for amendment to the Salem Unit 2 operating license is not required.

Document Control Desk 4

• MAY 151995 LR-N95077 The reload core design has been verified during the startup physics testing program. The program included the following tests:

1. Critical boron concentration measurements

2. Control rod bank worth measurements

3. Moderator temperature coefficient measurements

4. Power distribution measurements using the incore flux mapping system All measurements and test results were determined to be within their respective acceptance criteria. Test results are summarized in Attachment 1.

Should you have any questions regarding this transmittal, we will be pleased to discuss them with you.

Sincerely, Attachment (1)

Document Control Desk 5

• MAY 1 51995 LR-N95077

References:

1) Davidson, s. L. (Ed.), et. al., "Westinghouse Reload Safety Evaluation Methodology," WCAP-9273-NP-A, July 1985.

2) Salem Generating Station Updated Final Safety Analysis Report, USNRC Docket Numbers 50-272 and 50-311, as amended through Revision June 12, 1994.

3) Haessler, R. L., et. al., "Methodology for the Analysis of the Dropped Rod Event," WCAP-11394-A, January 1990.

4) Amendment No. 133 to License No. DPR-75, Boron Concentration in Boric Acid Tanks Reduced, Salem Nuclear Generating Station Unit 2 {TAC No. M86724), July 20, 1994.

5) Salem ,Generating Station Technical Specifications Unit 2, USNRC Docket Number 50-311, through Amendment No. 132, July 20, 1994. .

• 6) NLR-N94222,"Cycle 9 Radial Peaking Factor Limit Report, Salem Generating Station, Unit No. 2, Docket No. 50-311," Letter from s.

LaBruna to United States Nuclear Regulatory Commission, dated December 15, 1994.

Document Control Desk 6

• MAY 1 51995 LR-N95077 C Mr. T. T. Martin, Administrator - Region I U. s. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. L. N. Olshan, Licensing Project Manager - Salem U. s. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. C. S. 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

LR-N95077 Attachment 1 Salem Unit 2 Cycle 9 Startup Physics Test Results

1. Critical boron concentration measurements:

Measured Value Design Value Acceptance Sat/Unsat (ppm) (ppm) Criteria 1648 1688 +/- 4% Sat (1620 - 1756)

2. Control rod bank worth measurements:

Dilution Mode Rod Worth Results Rod Bank Measured Design Value Acceptance Sat/Unsat Worth (pcm) (pcm) Criteria D 922.9 857 +/- 15% Sat c 1209.8 1146 +/- 15% Sat B 1237.0 1181 +/- 15% Sat A 689.4 747 +/- 15% Sat SD 599 614 +/- 100 pcm Sat SC 715.5 709 +/- 15% Sat SB 1243.7 1210 +/- 15% Sat SA 104.0 96 +/- 15% Sat Total Banks 6747.8 6560 +/- 10% Sat

3. Moderator temperature coefficient measurements:

Measured Value Design Value Required Value Sat/Unsat (pcm/°F) (pcm/"F) (pcm/ °F)

-3.18 -2.4 < 0 Sat Page 1 of 2

LR-N95077 Attachment 1 Salem Unit 2 Cycle 9 Startup Physics Test Results

4. Power distribution measurements:

Parameter Test Measured Required Values Sat/Unsat Measured Conditions Values 1.55[1.0+0.3(1-P)]

(Power)

FNAH 27.4% 1.5684 < 1.8876 Sat Nuclear Enthalpy Hot 44.9% 1. 5531 < 1. 8062 Sat Channel Factor 94.6% 1. 5102 < 1. 5751 Sat Parameter Test Measured Required Sat/Unsat Measured Conditions Values Values (Power)

FQ (Z) 27.4% 2.0818 [4. 64] [K (Z) ] Sat Heat Flux < 4.6400 Hot Channel Factor 44.9% 1. 9976 [4. 64] [K (Z) ] Sat

< 4.6400 94.6% 1.8487 [2.32] [K (Z)] Sat

< 2.4524 Parameter Test Measured Required Sat/Unsat Measured Conditions Values Values (Power) c 27.4% Rodded= 1.7454 Rodded< 2.6061 Sat F XY Computed Radial Unrodded = 1.6605 Unrodded <

Peaking 2.0581 Factor 44.9% Rodded = 1. 8 2 77 Rodded < 2.4937 Sat Unrodded = 1. 6387 Unrodded <

1.9694 94.6% Rodded = N/A Rodded < N/A Sat Unrodded = 1.6034 Unrodded <

1.7174 Page 2 of 2