ML20044H412
| ML20044H412 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/28/1993 |
| From: | George Thomas DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20044H413 | List: |
| References | |
| NUDOCS 9306080364 | |
| Download: ML20044H412 (2) | |
Text
'.
8 Bea er Vaney Power Station SNppmgport. PA 15077-0004 (412) 393-5206 (412) 643-8069 F AX GEORGE $ THOMAS Dmsion Vme Pres < dent swear semm May 28, 1993 Nuclear Power Dmuson i
U.
S.
Nuclear Regulatory Commission Attn:
Document Control Desk Washington, DC 20555
Subject:
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Cycle 10 Reload and Core Operating Limits Report Beaver Valley Power Station, Unit No. 1 completed the ninth cycle of operation on March 26, 1993, with a burnup of 18,339 MWD /MTU.
This letter describes the Cycle 10 reload design, documents our review in accordance with 10 CFR 50.59 and our determination that no technical specification changes or unreviewed safety questions are involved, and provides a
copy of the Core Operating Limits Report (COLR) in accordance with Technical Specification 6.9.1.14.
The new core configuration is arranged in a low leakage loading pattern and involves removing one (1) Region 1, twelve (12) Region 9A,.
sixteen (16)
Region 10A and fifty-two (52)
Regicn 10B fuel assemblies.
These are replaced with sixty-four (64) fresh Region 12 fuel assemblies enriched to 3.6 weight percent along with one (1)
Region 1,
eight (8)
Region 9A and eight (8)
Region 9B fuel assemblies.
The mechanical design of the new Region 12 fuel assemblies is the same as the previous cycle reload assemblies.
Fuel rod design evaluations for the new fuel were performed using NRC l
approved methodology to demonstrate that all of the fuel rod design bases are satisfied.
Duquesne Light Company has performed a detailed review of this reload core design including a review of the core characteristics to determine those parameters affecting the postulated accidents described in the Updated Final Safety Analysis Report (UFSAR).
The consequences of those incidents described in the UFSAR which could i
potentially be affected by the reload core characteristics were evaluated in accordance with the NRC approved methodology described in WCAP-9272-P-A " Westinghouse Reload Safety Evaluation Methodology."
The effects of the reload design can be accommodated within the conservatisms of the assumptions used in the current analysis design
- basis, or it was demonstrated through evaluation that the reload parameters would not change the conclusions in the UFSAR.
9 0 (I}
I 93060B0364 930528 PDR ADDCK 05000334 L
p PDR
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Cycle 10 Reload and Core Operating Limits Report Page 2
't I
No technical specification changes are required as a result.of this reload design.
The NRC approved dropped rod methodology
[WCAP-10298-A (non-proprietary),
June 1983] was used for this design evaluation and confirmed that the peaking factors did not exceed the safety analyses limits.
The reload core design will be verified by performing the standard Westinghouse reload core physics startup tests.
The results t
of the following startup tests will be submitted in accordance with Technical Specification 6.9.1.3:
1.
Control rod drive tests and rod drop time measurements.
2.
Critical boron concentration measurements.
3.
Control rod bank worth measurements.
4.
Moderator temperature coefficient measurements.
5.
Startup power distribution measurements-using the incore flux mapping system.
The COLR (attached) has been updated for this reload to include new F
(RTP) limits for unrodded core planes and Figure 4 has been replaced with a new figure to address these new limits.
x The Beaver Valley Onsite Safety Committee (OSC) and the Duquesne Light Company Offsite Review Committee (ORC) have reviewed the Reload Safety Evaluation and Core Operating Limits Report and determined that this reload design will not adversely affect the safety of the plant and does not involve an unreviewed safety question.
Sincerely,
- ,cic D
- c3 G.
S.
Thomas cc:
Mr.
L.
W.
Rossbach, br. Resident Inspector Mr.
T.
T.
Martin, NRC Region I Administrator Mr.
G.
E.
Edison, Project Manager Mr.
M.
L.
Bowling (VEPCO) 1