ML20214D691
| ML20214D691 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 05/06/1987 |
| From: | Ainger K COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8705210431 | |
| Download: ML20214D691 (4) | |
Text
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V-Chca00.Isnois80000 0767 May 6, 1987 U.S. Nuclear Regulatory Commission Attn: -Document Control Desk Washington, DC= 20555
Subject:
_ Byron Station Unit 1 Cycle 2 Reload Etc Docket No. 50-454 References (a): Westinghouse WCAp-10021-NP-A, October 1983'
" Westinghouse Wet Annular Burnable Absorber Evaluation Report"
'(b): July 27, 1983 letter from F.G. 1.entine to H.R.
Denton " Commonwealth Edison Topical Report on Benchmark of PWR Nuclear Design Methods" (c): NRC Safety Evaluation Report on Neutronics Topical (Ref. (b)) December 13, 1983 (d): January 6, 1987 letter from S.C. Hunsader to H.R. Denton Gentlemen:
Byron Unit I has recently completed its first cycle of operation and is near the end of a refueling outage which began on February 14,1987. Cycle 1 terminated with'a cycle burnup of 17,937 MWD /MTU. Startup for Cycle 2 is expected to occur in mid-May, 1987. This letter is to advise you of Commonwealth Edison Company's plans regarding the Byron Unit 1 Cycle 2 reload core and the conclusions of our review per 10 CFR 50.59.
The Byron Unit 1, Cycle 2 reload core was designed to perform under current nominal design parameters, Technical Specifications and related bases, and current setpoints such that:
1.
Core characteristics will be less limiting than those previously reviewed and accepted; or 2.
For those postulated incidents analyzed and Ieported in the Byron Final Safety Analysis Report (FSAR) which could potentially be affected by fuel reload, reanalysis has demonstrated that the results of the postulated events are within allowable limits. Commonwealth Edison Company performed a detailed review with Westinghouse on the bases, including all the postulated incidents considered in the PSAR, of the Westinghouse Reload Safety Evaluation (RSE).
Based on this review and the Westinghouse RSR, safety evaluations were performed by Commonwealth Edison on-Site and off-Site Review pursuant to the requirements of 10 CFR 50.59(a) and 10 CPR 50.59(b).
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. The Byron Unit 1 core is being refueled with Westinghouse's 17x17 optimized fuel assemblies (OFA). Although the reload fuel mechanical, nuclear, and thermal-hydraulic design for the Cycle 2 reload core has not significantly
. changed:from that of the previously reviewed and accepted initial design, this cycle introduces the first Byron application of Westinghouse's Wet Annular Burnable Absorbers (reference (a)), chamfered fuel pellets and reduced rod bow top and middle grids. Additionally, this reload incorporates one reconstituted fuel assembly which contains two stainless steel rods.
The current FNDH limit of less than 1,55 ensures that the DNB ratio remains above 1.49 for a typical fuel cell and 1.47 for a thimble cell.
In addition, based upon the performance of an eighteen case FAC analysis, a total peaking factor (F ) of.2.314 is the maximum which could occur for the full q
range of power distributions, including load follow maneuvers allowable under constant axial offset control (CAOC). The current. radial peaking factor j
(Fgy) limits of 1.55 for unrodded core planes and 1.71 for core planes i-containing Bank "D" control rods will be used to ensure that Fq remains i
within its limits. Therefore,' additional surveillance of Fq(2) is not required in cycle 2 for compliance with the approved Byron Unit 1 and 2 Fq j-limit of 2.32.
1 Post-Cycle 1 ultrasound examinations of all Byron Unit I fuel identified four fuel assemblies which contained leaking fuel rods. All four l
assemblies were reconstituted by replacing the leaking fuel rods with non-fuel i
filler rods fabricated from stainless steel. Three of the reconstituted assemblies were discharged (i.e., remain in spent fuel storage facility).
Assembly B34 containing two stainless steel rods (assembly coordinates C-10 and Q-11) was reinserted into the cycle 2 core (location E-09).
has evaluated the effect of two stainless steel rods on the non-LOCA safety analyses for cycle 2 and concluded that there is no impact.
For the 14CA, Westinghouse has concluded that the use of two stainless steel rods in cycle 2 will result in a peak cladding temperature (PCT) change of less than l'F.
i Substantial margin ~to the 2200*F limit is available to absorb this penalty based on the analysis of record (maximum PCT of 2110*F).
A The new fuel regions utilize 1) fuel pellets which have a small chaefer at the outer edge of the pellet ends and a reduction in the dish diameter and depth compared to previous chaefered pellets, 2) top grids with a j
reduced spring force and middle grids with a redesigned spring to reduce potential rod bow, 3) 4g fuel rod plenum springs, and 4) reduced fuel rod backfill pressure. The impact of these minor enhancements has been evaluated and found to be acceptable by Westinghouse and Commonwealth Edison.
It should be noted that analyses have also been performed by Westinghouse which justify operations at a hot leg temperature (Thot) below l
t the PSAR reported nominal 618'F in order to minimize potential primary water
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stress corrosion cracking in the steam generator tubes. Although Commonwealth Edison does not currently plan to implement a Thot reduction program for the j
Cycle 2 startup, the new analyses (including LOCA) will be submitted for NRC review and approval prior to implementation of Thot reduction during Cycle 2.
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U.Si NRC _May 6, 1987 As in the past, the reload safety evaluation relied on previously reviewed and accepted analyses reported in the PSAR and fuel densification reports. A detailed review of the core characteristics was performed to determine those parameters affecting the analyses of postulated accidents reported in the Byron /Braidwood FSAR.
For those incidents whose consequences could potentially be affected by the reload core characteristics, the incidents were reanalyzed. r==anwealth Edison verified that the reanalyses were performed in accordance with Westinghouse reload safety evaluation methodology, as outlines in the March 1978 Westinghouse Topical Report entitled
" Westinghouse Reload Safety Evaluation Methodology" (WCAP-9272), and were consistent with the PWR neutronic methods currently employed by Cosmonwealth Edison, as qualified in the reference (b) topical report and related NRC SER in reference (c). Commonwealth Edison also verified that the results of these reanalyses were within previously reviewed and accepted limits.
The reload safety evaluation demonstrated that core related Technical Specification changes are not required for operation of Byron Unit I during Cycle 2, other than the previously submitted amendment request regarding fuel assembly _ reconstitution (reference (d)). Commonwealth Edison on-Site and off-site Reviews concluded that no unreviewed safety questions, as defined by 10 CFR 50.59, are involved with this reload. More specifically, with this reload:
1.
There is no increase in the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in'the safety analysis report; 2.
No additional accident or malfunction of a different type than any evaluated previously in the safety analysis reported has been created; and 3.
There has been no reduction in the margin of safety as defined in the basis for any Byron Unit 1 Technical Specification.
Therefore, prior NRC review and approval of the reload core analyses and application for amendment to the Byron Unit 1 operating license is not required as a result of the cycle specific reload analyses for Cycle 2.
Finally, verification of the Byron reload core design will be accomplished per the standard PWR startup physics tests normally performed at the start of each reload cycle and as identified in ANSI /ANS-19.6.1-1985.
These tests include, but are not limited to:
1.
A physical inventory of the fuel in the reactor by serial number and location prior to the replacement of the reactor head; 2.
Control rod drive tests and drop times;
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' May 6, 1987 3.
Critical boron concentration measurements; 4.
Control bank worth measurements using the rod swap technique; s'
5.
Moderator temperature coefficient measurements; and 6.
Startup power distribution measurements using the incore flux mapping system.
If you have any questions regarding this matter, please contact this office.
Very truly yours, i
K. A. Ainger Nuclear Licensing Administrator 1m cct Byron Resident Inspector NRC Region III Office 3039K