ML20211Q604
| ML20211Q604 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/31/1999 |
| From: | Sorensen J NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20211Q607 | List: |
| References | |
| NUDOCS 9909150103 | |
| Download: ML20211Q604 (4) | |
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eff Northern States Power Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch, Minnesota 55089 August 31,1999 10 CFR 50.71(e)
U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Submittal of Revision No.19 to the Updated Safety Analysis Report (USAR)
Pursuant to 10 CFR 50.71(e) we are submitting one original and 10 copies of Revision No.19 to the Updated Safety Analysis Report (USAR) for the Prairie Island Nuclear Generating Plant. This revision brings the USAR up-to-date as of February 28,1999, though some information is more recent. contains descriptions and summaries of safety evaluationt for changes, tests, and experiments made under the provisions of 10 CFR 50.59 during the period since the last update. contains the USAR page changes and instructions for entering the pages.
In this letter we have made no new Nuclear Regulstory Commission commitments. kC 6
I certify that the information presented herein accurately presents changes made since the last updating submittal of the Prairie Island USAR.
9909150103 990831 PDR ADOCK 05000282 l
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USNRC August 31,1999 i
Please contact Jack Leveille (651-388-1121, Ext. 4124) if you have any questions related to this letter.
Joel P. Sorensen Site General Manager Prairie island Nuclear Generating Plant c: Regional Administrator-Region lil, NRC Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg Attachments: 1. Safety Evaluation Summaries
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Attachm:nt 1 Page 1 of 2 ATTACHMENT 1 l
PRAIRIE ISLAND NUCLEAR GENERATING PLANT REPORT OF CHANGES, TESTS AND EXPERIMENTS - AUGUST 1999 l
Below are a brief description and a summary of the safety evaluation for each of those changes, tests, and experiments which were carried out without prior NRC approval, l
pursuant to the requirements of 10 CFR Part 50, Section 50.59(b).
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i Modification 99FH01 - Unit 1 Cycle 20 Reload i
Description of Chanae This modification replaced depleted Unit 1 fuel assemblies with a fresh reload of 48 Westinghouse VANTAGE + fuel assemblies allowing another cycle of power operation. All l
of the new assemblies are enriched to a nominal 4.95 w/o U235 and result in a projected cycle length of 23302 mwd /MTU.
The Unit 1 Cycle 20 reload was developed by NSP Nuclear Analysis & Design (NSPNAD) using approved methodology addressed in NSPNAD-8101-A, Qualifications of Reactor Physics Methods for Application to PI Units. More details on the operational parameters l
can be found in NSPNAD-99002P, Rev. O, Prairie Island Unit 1 Cycle 20 Startup and l
Operations Report, and NSPNAD-99001, Rev. O, Prairie Island Unit 1 Cycle 20 Final Reload Design Report and USAR Update.
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l Summarv of Safety Evaluation The following safety concems were addressed in the safety evaluation:
A.
Thermal Hydraulic Analysis l
B.
Accident and Transient Analysis l
C.
Uncontrolled Boron Dilution D.
Main Steam Line Break / Containment Response Analysis E.
LOCA-ECCS Analysis F.
Rod Ejection Analysis l
G.
Fuel Handling Accident l
H.
Refueling Shutdown Margin l.
Heatup/Cooldown Curves - Reactor Vessel Radiation Surveillance Program J.
Fuel Rod Design Performance K.
Spent Fuel Pool Heat Load L.
New Fuel Rack / Spent Fuel Rack Criticality M.
Core Exposure Limits /Off-site Dose Calculations N.
Peak Linear Heat Generation Rate O.
Fuel Assembly Design Change P.
Startup and Operations
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Att:chmint 1 Page 2 of 2
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Validity of Safety Evaluation All results were accsptable and are presented in NSPNAD-99001, Rev. O, Prairie Island Unit 1 Cycle 20 Final Reload Design Report and USAR Update. The LOCA analysis was performed by Westinghouse and is documented in the Unit 1 Cycle 20 LOCA Confirmation Letter 99NS-G-0009, April 6,1999. This letter confirms that Unit 1 Cycle 20 will continue to conform to the acceptance criteria of 10CFR50.46.
Since all transient ana!yses meet the acceptance criteria, there are no unreviewed safety questions for the Unit 1 Cycle 20 Core Reload Design Change.
License Amendments 144 and 135-Incorporation of Combustion Engineering Steam Generator Welded Tube Sleeve Topical Report Revision Description of Chance The amendments changed certain requirements for repair of defective steam generator tubes based on the latest revision to a previously approved methodology.
Summary of Safety Evaluation The License Amendments were issued April 15,1999.
License Amendments 145 and 136 - Delay implementation of Additional License Condition No. 6 Description of Chanae The amendments change the implementation due date for Condition No. 6 (relocation of the requirements specified in technical specification Sections 3.1.E and 5.1 to the Updated Final Safety Analysis Report) from June 1,1999, to September 1,1999.
Summarv of Safety Evaluation The License Amendments were issued June 2,1999.
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