ML20195C686

Forwards Rev 17 to USAR for Prairie Island Nuclear Generating Plant.Attachment 1 Contains Descriptions & Summaries of SEs for Changes,Tests & Experiments Made Under Provisions of 10CFR50.59 During Period Since Last Update

ML20195C686
Person / Time
Site:	Prairie Island
Issue date:	05/21/1999
From:	Sorensen J NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20195C691	List: ML20195C686
References
NUDOCS 9906080272
Download: ML20195C686 (10)

Text

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.p Northern States Power Company i

Prairie island Nuclear Generating Plant 1717 Wakonade Dr. Esst Welch, Minnesota 55089 May ?1,1999 10 CFR 50.71(e)

U S Nuclear Regulatory Commission Attn: Document Control Desk ,

Washington, DC 20555 I l

L PRAIRIE ISLAND NUCLEAR GENERATING PLANT I Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Submittal of Revision No.17 to the I' Updated Safety Analysis Report (USAR)

Pursuant to 10 CFR 50.71(e) we are submitting one original and 10 copies of Revision No.17 to the Updated Safety Analysis Report (USAR) for the Prairie Island Nuclear Generating Plant. This revision brings the USAR up-to-date as of December 31,1998 (though some information is more recent) with one exception:

proposed changes identified by the USAR Review Project Team are not included in this revision and will be submitted shortly as a separate revision.

Attachment 1 contains descriptions and summaries of safety evaluations for changes, tests, and experiments made under the provisions of 10 CFR 50.59 during the period since the last update. Attachment 1 also contains discussions of changes made to regulatory commitments made within our Regulatory Commitment Change Process.

Attachment 2 contains the USAR page changes and instructions for entering the pages.

In this letter we have made no new Nuclear Regulatory Commission commitments.

I certify that the information presented herein accurately presents changes made since the last updating submittal of the Prairie Island USAR.

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1 USNRC May 21,1999 Please contact Jack Leveille (651-388-1121, Ext. 4142) if you have any questions related to this letter. ,

l Joel P. Sorensen  !

Site General Manager l Prairie Island Nuclear Generating Plant c: Regional Administrator - Region Ill, NRC Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg Attachments: 1. Safety Evaluation Summaries

2. USAR page changes

ATTACHMENT 1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT  ;

REPORT OF CHANGES, TESTS AND EXPERIMENTS - MAY 1999 The following sections include a brief descriptisn and a summary of the safety evaluation for each of those changes, tests, and experiments which were carried out without prior NRC approval, pursuant to the requirements of 10 CFR Part 50, Section 50.59(b). Also included are discussions of changes made to regulatory commitments made within our Regulatory Commitment Change Process ,

Modification 96 COO 6 Part A- Steam Release Computer Replacement Description of Chance This modification to replace the Steam Release Computer will be performed in two parts. This part, Part A, moved the calculations performed by the present Steam l Release Computer and all required field inputs to the ERCS computer. Part B of the modification will remove the old Steam Release Computer and the Recall Computer, and all associated wiring will be removed or abandoned.

i Summary of Safety Evaluation This modification affects a calculational tool used in the determination of offsite radioactive release concentrations via steam releas.e headers. As such, the effects of this modification were reviewed against the Emergency Plan 50.54(q) screening form to determine if prior NRC approval of the change is required. The results of this review indicate that, although a change is being made to the equipment used for l assessing consequences of offsite radiological release, the change does not decrease the effectiveness of the Emergency Plan. The move of the calculation from  !

the dedicated Steam Release Computer to the ERCS computer results in a calculation equivalent to the original Steam Release Computer results.

I Modification 97AF02 - AMSAC/ Diverse Scram System Description of Chanae This safety evaluation addresses the additional changes to the feedwater control system and the AMSAC/ DSS system not specifically addressed in SER dated September 22,1998. The additional changes include the upgrade of the data highway communications cards in AMSAC and Feedwater systems, the changeout of engineering workstations in the Feedwater system, and the changeout of the I/O card cage and input and output cards in the AMSAC/ DSS cabinet.

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May 21,1999 S'ummarv of Safety Evaluation I

The changeout'and upgrade of hardware in the AMSAC and Feedwater systems was necessitated by obsolescence of existing hardware. The functionality of the -

replacement hardware is equivalent to the presently installed equipment. Installation activities for this modification occur during a refueling outage, while the plant is in Cold Shutdown or Refueling Mode. In this mode, the feedwater control system, the AMSAC system, and the rod control system are not in service. Following installation, the changes were testod using preoperational tests in Cold Shutdown or Refueling .

mode. The Auxiliary Feedwater System, while the motor-driven pump is still operable in this mode for cross-connect to the other unit, is not affected by AMSAC/ DSS when the motor-driven pump mode selector switch is placed in

' Shutdown Auto' or ' Manual'. The activities did not cause any system to be operated outside of its design or testing limits and, therefore, the activities did not result in an unreviewed safety question. An additional preoperational test will be performed in hot _ shutdown with the plant in a condition which allows testing coincident with rod

! drop testing. The Auxiliary Feedwater Pumps are required to be operable in hot shutdown, and this preoperational test maintains that operability. No LCOs are entered during the performance of this test.

License Amenoments 138 and 129 authorized modification to AMSAC.

l Modification 98FH01 - Unit 2 Cycle 19 Reload Description of Chance j This modification replaced depleted Unit 2 fuel assemblies with a fresh reload of 45 Westinghouse VANTAGE + fuel assemblies allowing another cycle of power operation.

Forty-four of the new assemblies are enriched to a nominal 4.95 w/o U235 and one is i enriched to 3.80 w/o U235 (center bundle) and results in a projected cycle length of 18779 mwd /MTU.

l The Unit 2 Cycle 19 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 ttie operational parameters can be found in NSPNAD-98005, Rev. O, Prairie Island Unit 2 Cycle 19 Startup and Operations Report, and NSPNAD-98004, Rev.1, Prairie L island Unit 2 Cycle 19 Final Reload Design Report.

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Page 3 of 7 May 21,1999 Summary of Safety Evaluation The following safety concerns were addressed in the safety evaluation:

A. Thermal Hydraulic Analysis 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 G. Fuel Handling Accident H. Refueling Shutdown Margin I. 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 Q. Validity of Safety Evaluation All results were acceptable and are presented in NSPNAD-98004, Rev.1, Prairie Island Unit 2 Cycle 19 Final Reload Design Report. The LOCA analysis was performed by Westinghouse and is documented in the Unit 2 Cycle 19 LOCA l Confirmation Letter 98NS-G-0018, July 31,1998. This letter confirms that Unit 2 ,

Cycle 19 will continue to conform to the acceptance criteria of 10CFR50.46. l Since all transient analyses meet the acceptance criteria, there are no unreviewed safety questions for the Unit 2 Cycle 19 Core Reload Design Change.

Safety Evaluation 340, Rev.1 - Fan Coil Unit Damper Control Circuit Configuration Description of Chance The purpose of this evaluation is to review the control wiring circuitry for the Containment Fan Coil Unit (FCU) dampers. The wiring is currently routed as nonsafety-related. This evaluation documents that the current configuration cannot defeat the safety-related function of the FCU dampers.

Summary of Safety Evaluation This safety evaluation creates no unreviewed safety questions. This is based on the following:

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May 21,1999 This evaluation concludes that the postulated failure mechanism which could cause all of the FCU discharge dampers to align to the nonsafeguards position is not '

credible. Since the failure mechanism is not credible, the post-accident function of the FCU's is maintained. As the FCU's can complete their post-accident function (containment heat removal), the containment pressure analysis is not affected.

Therefore, there is no potential increase in consequences or reduction in the margin.

of safety. The FCU's and associated dampers are accident mitigation equipment; they cannot initiate an accident of a different type or increase the probability of an accident occurring. As the postulated failure mode is not credible, there is no new type of equipment malfunction or increase in the probability of an equipment malfunction.

Safety Evaluation 520 - Evaluation of Control Room Habitability During a Postulated Onsite Hazardous Chemical Release Description of Chanae The safety evaluation documents the evaluation of onsite chemicals for their effects on control room habitability.

Summary of Safety Evaluation Onsite chemicals are evaluated on a periodic basis. The evaluation concludes that control room operator incapacitation will not occur following a postulated hazardous onsite chemical release. The calculations and review are based on the chemicals' toxicity and properties that effect mass release.

Safety Evaluation 527-11 Changes to USAR Section 11.9.4.2, " Wall Thickness Monitoring of High Energy Piping" Description of Chance USAR Section 11.9.4.2,"WallThickness Monitoring of High Energy Piping" describes Prairie Island's inspection program for pipe wall thinning. Statements regarding selection of inspection points and methods of thickness measurement were found to be no longer accurate. As experience was gained and technology i

improved, inspection methods were improved; however, the USAR was not revised to reflect the new practices. The revision to this section removes some detail and t

' corrects the inaccuracies.

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! May 21,1999 Summary of Safety Evaluation This USAR change does not alter any system, structure or component function.

Operation and design of all systems remains unchanged. Since the changes are )

only to the inspection program, consequences of accidents or malfunctions cannot  !

be affected, nor can new accidents or malfunctions be created. The inspection  !

program can only affect the probability of accidents or malfunctions (pipe breaks due to pipe wall thinning). Since these changes will make the Pipe Wall Thinning program more effective, the probability of accidents or malfunctions will not be i

increased. Also, since these changes have no effect on the results of any analyses, I 4

there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

Safety Evaluation 531 - Changes to USAR Section 12.2.7, Turbine Missiles Description of Chanae i Low pressure turbine rotors utilizing shrunk-on discs were removed from service and ,

replaced with fully integral nuclear LP rotors. Newer forging technology has allowed j the rotor to be a single forging and therefore eliminated the disc keyways and bores. l Rotor peak stresses are significantly reduced, leading to a large reduction in probability of a rotor burst.

< Summary of Safety Evaluation l Following installation of the fully integral LP rotors, the probabilities of missile ejection at running and design overspeed are less than 5X104 even after 20 years of running time it is concluded that periodic inservice inspections are not required for the new rotors.

License Amendments 138 and 129 - Modification to ATWS Mitigating System Actuating Circuitry Description of Chanae The amendments authorize a design modification of the existing Anticipated Transient Without Scram (ATWS) Mitigation System Actuati0on Circuitry (AMSAC).

The design modification would install a Diverse Scram System (DSS) designed to meet the requirements ,of a DSS described by 10CFR50.62 and make major modifications to the existing AMSAC.

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May 21,1999 Sumii,ary of Safety Evaluation

-The amendments were issued September 22,1998.

I License A6nendments 141 and 132- Revised Administrative Controls '

Description of Chanoe The amendments revise TS 6.0, Administrative Controls, and the following TS sections affected by relocating, removing, and modifying requirements to TS 6.0:

Table' of Contents; TS 3.1, " Reactor Coolant System"; TS 4.0, " Surveillance i Requirements"; TS 5.0," Design Features"; and associated Bases. The mmoved or l relocated requirements are adequately controlled by existing regulations other than 10CFR50.36 and the TS.

Summary of Safety Evaluation i The License Amendments were issued December 7,1998, though full implementation may not take place until September 1,1999.

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e a Attachment 1 Page 7 of 7 May 21,1999 CHANGES TO REGULATORY COMMITMENTS Regulatory Commitment Change 99-01 Corrective action stated in Unit 1 LER 98-012 was to " Review corr.pliance with exemptions to 10CFR50 Appendix R for all Fire Areas." Due date was March 31, 1999. Review of exemptions with respect to each previous submittal of the Safe Shutdown Analysis is more complex and is taking longer than anticipated. The due date was changed to April 30,1999.

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<.d' ATTACHMENT 2 PRAIRIE ISLAND NUCLEAR GE NERATING PLANT Revision 17 to the Updated Safety Analysis Report ,

Instructions:

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1. Remove and discard individual USAR pages, tables, and figures and replace with the new Revision 17 pages provided. Specialinstructions, where applicable, are included with the replacement pages.

2. When page removal / replacement is complete, review the USAR List of Effective Pages to ensure your copy of the USAR is current and complete. Contact NSP Nuclear Licensing at 651-388-1121, Extension 4142 if you require additional assistance. I i

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