ML20203E922

Forwards Rev 4 to Updated FSAR for Prairie Island Nuclear Generating Plant.Two New Vols Provided for Reprinting of Apps of Original FSAR

ML20203E922
Person / Time
Site:	Prairie Island
Issue date:	06/30/1986
From:	Musolf D NORTHERN STATES POWER CO.
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20203E926	List: ML20203E922
References
TAC-61635, NUDOCS 8607300021
Download: ML20203E922 (5)

Text

.o -

)

Northern States Power Company 414 Nicollet Mall Minneapoks. Minnesota 55401 Telephone (612) 330-5500 June 30, 1986 Director Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Submittal of Revision No. 4 to the Updated Safety Analysis Report (USAR)

Pursuant to 10CFR 50.71(e) we are submitting 14 copies of Revision No. 4 to the Updated Safety Analysis Report (USAR) for the Frairie Island Generating Plant. This revision updates the information in the USAR for the period from January 1, 1985 through December 31, 1985.

Exhibit A contains a description and a summary of the safety evaluations for changes, tests and experiments made under the provisions of 10CFR 50.59 during this period.

Exhibit B contains the USAR page changes and instructions for entering the pages.

This revision also includes a reprinting of the appendices of the original Final Safety Ana.tysis Report. Two new volumes are provided for this purpose.

There have been no changes to the Northern States Power Company Operational Quality Assurance Plan, the Licensed Operator Training Program, or the Licensed Operator Requalification Program. The Licensed Operator Training Program, the Licensed Operator Requalification Program and the Operational Quality Assurance Plan have been relocated from the Chapter 13 appendices to Appendix A, Appendix B and Appendix C respectively.

D ld > A David Musolf Manager - Nuclear Support Services DMM/EFE/efe c: Regional Administrator-III, NRC Director IE, NRC (w/o Exhibit B) /

NRR Project Manager, NRC (w/o Exhibit B)

Resident Inspector, NRC (w/o Exhibit B)

C Charnoff (w/o Exhibit B)

I

\

j\ Y Attachments 86973 0 860630

m. .05000282, PDR J

. o EXHIBIT A PRAIRIE ISIAND NUCLEAR GENERATING PLANT ANNUAL REPORT OF CHANGES, TESTS AND EXPERIMENTS January 1, 1985 to December 31, 1985 The following sections include a brief description and a summary of the safety evaluation for those changes, tests and experiments which were carried out without prior NRC approval, pursuant to the requirements of 10CFR 50.59(b).

1. CRACKS IN CONTROL ROD TIPS (SE #153)

Description of Change Control Rods at Prairie Island have been discovered to have small hairline cracks on some rodlet tips. The cracks are caused by Intergranular Stress Corrosion Cracking of the stainless steel cladding enhanced by irradiation swelling of the absorber material as well as cladding thickness wear.

Summary of Safety Evaluation EPRI hot cell examinations of similar control rod cracks have shown no significant safety concerns. The cracks are too small to affect control rod drop times or to allow significant leaching of the absorber material out of the rods. The cracks are axially oriented and as such do not pose any loose parts concerns.

2. REACTOR VESSEL LEVEL INSTRUMENTATION SYSTEM (80Y124)

Description of Change This design change provides for the installation of two redundant trains of reactor vessel instrumentation for each unit. For each train, the reactor vessel is tapped at three elevations and impulse lines transmit pressure to three differential pressure transmitters located in the Auxiliary Building. For each train, a microprocessor reads the data from the differential pressure transmitters, as well as other pertinent data provided to the microprocessor, and then calculates a compensated ranctor vocac1 water inventory. The calculated level in then transmitted to a display in the Control Room.

Summary of Safety Evaluation All equipment necessary for the operation of the reactor vessel level instrumentation system meets the requirements of NUREG-0737, Item I I , F. 2. The instruments, microprocessors and display modules are tested in accordance with IEEE-323-1974 and IEEE-344-1975. Also, as a part of the plant's upgrade of instrumentation to the requirements of Regulatory Gulde 1.97, the reactor vessel level instrumentation system's power supply is being upgradet'. to a qualified Class lE system.

A-1

l

3. NO. 11 STATION BATTERY REPLACEMENT (84YS00)

Description of Change This design change replaced No.11 Station Battery with a newer and larger battery. The replacement of the battery rack required relocation of the Triaxial Peak Recording Accelerograph located in No. 11 Battery Room.

Sumary of Safety Evaluation The replacement battery for No. 11 Station Battery is larger than the original battery and will therefore exceed the USAR requirements.

Analysis shows that the battery charger is large enough to charge the larger battery within the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> required by the USAR. Replacement of the battery was performed while Unit 1 was shutdown for refueling and common loads required for Unit 2 operation were transferred to the Unit 2 batteries. The seismic peak recording accelerograph was relocated close to its original location in No. 11 Battery Room. Therefore this design change does not present an unreviewed safety question.

4. REACTOR TRIP BREAKER SHUNT TRIP AUTOMATIC ACTUATION (84Y545)

Description of Change As a result of this modification automatic reactor trip signals now energize the shunt trip coils on the DB-50 reactor trip breakers (while deenergizing the undervoltage trip assemblies). The modification was accomplished by creating a relay trip logic matrix on spare contacts of existing reactor trip logic relays.

Summary of Safety Evaluation The ability to monitor reactor trip breaker status is not diminished.

System response to a seismic event will not inhibit reactor trip. No inadvertent reactor trip resulting from the single failure of a device is introduced by this modification. Therefore this modification does not increase the consequences of an accident or reduce the plant's margin of safety.

5. REINSTATE THE PRESSURIZER HIGH/ LOW ALARM (84L826)

De_scription of Ch*.nge In response to a TMI Action Plan requirement to alarm an inadvertent PORV opening, the pressurizer low pressure alarm was changed from 2185 psig to 2335 psig. The logic for the high pressure alarm remained at 2310 psig and as a result the pressurizer High/ Low alarm was on at all times.

To make the circuit function correctly, the high alarm logic was changed to low alarm logic and the setpoint was changed. The low pressure alarm is set at 2185 psig and the high pressure setpoint remains at 2335 psig.

i Summary of Safety Evaluation This modification reactivated the Control Room pressurizer High/ Low alarm indicator which had been out of service. The low pressurizer pressure setpoint was set at 2185 psig as per the original plant design.

Therefore this modification does not involve an unreviewed safety question.

l A-2

(

s =

6. UNIT 1 CYCLE 10 CORE RELOAD (84L847)

Description of Change This design change addresses the Core Reload for Unit 1 Cycle 10.

During the Prairie Island Unit 1 Cycle 9/10 refueling outage, 40 spent Exxon fuel assemblies and 1 Westinghouse fuel assembly were replaced with 40 fresh Exxon TOPROD assemblies and 1 twice burned Westinghouse assembly. The remainder of the core consisted of 20 once burned Exxon TOPROD assemblies, 60 twice' burned Exxon TOPROD assemblies. Cycle 10 is expected to reach an end of cycle exposure of 11,223 MWD /MTU. Unit 1 Cycle 10 utilized the Rod Swap Method to determine control rod reactivity worth during Physics testing instead of the boron dilution method.

Summary of Safety Evaluation The analyses performed in the design and licensing of Unit 1 Cycle 10 operation were done by NSP's Nuclear Analysis Department (NAD) and are summarized in the " Prairie Island Unit 1, Cycle 10 Final Reload Design Report (Reload Safety Evaluation)", October, 1984 and the " Prairie Island Unit 1, Cycle 10 Startup and Operations Report Rev 1", February, 1985. The analyses indicate that the core will operate within Technical Specification limits with respect to shutdown reactivity worth, temperature coefficient restrictions and hot channel peaking factors F delta H and FQ.

7. UNIT 2 CYCLE 10 CORE RELOAD (85L862)

Description of Change This design change addresses the Core Reload for Unit 2 Cycle 10.

During the Prr.irie Island Unit 2 Cycle 9/10 refueling outage, 57 Exxon TOPROD fuel assemblies were replaced. The reload for Unit 2 Cycle 10 consists of 40 new Exxon TOPROD assemblies, 40 once burned Exxon TOPROD assemblies, 16 twice burned Exxon TOPROD assemblies, 8 thrice burned TOPROD assemblies, 13 twice burned low enriched Exxon TOPROD assemblies and four twice burned Westinghouse standard assemblies. Cycle 10 is expected to reach an end of cycle exposure of 11,582 MWD /MTU. Unit 2 Cycle 10 utilized the Rod Swap Method to determine control rod reactivity worth during Physics testing instead of the boron dilution method.

Summary of Safety Evaluation The analyses performed in the design and licensing of Unit 2 Cycle 10 operation were done by NSP's Nuclear Analysis Department (NAD) and are summarized in the " Prairie Island Unit 2, Cycle 10 Final Reload Design Report (Reload Safety Evaluation) Rev 1", October 1985 and the " Prairie Island Unit 2, Cycle 10 Startup and Operations Report Rev 1", October 1985. The analyses indicate that the core will operate within Technical Specification limits with respect to shutdown reactivity worth, temperature coefficient restrictions and hot channel peaking factors F delta H and FQ.

A-3

( <

s ,e o

l'

8. IMPROVE RELIABILITY OF RADIATION MONITOR CHANNELS 1R-15 AND 2R-15 BY REPLACING EXISTING GM-TUBE DETECTORS (84L849)

Description of Change Replaced GM-tube with NaI detector and provided environmental barrier between sampling media and detector.

i Summary of Safety Evaluation The new R-15 scintillation detectors will perform the same function as the old GM detectors but with increased reliability because the detectors will not be immersed in the warm humid air. The range of the new detectors is equivalent to the original R-15 detectors. The effluent: stream will still be routed to the Auxiliary Building ventilation systems for further monitoring prior to release.

9. COMPONENT COOLING SURGE TANK RELIEF VALVE REMOVAL AND REPIACEMENT WITH SPOOL PIECE (84L850)

Description of Change The component cooling surge tank relief valve was removed and replaced with pipe and fittings on both units.

Summary of Safety Evaluation This change corrects an overpressure condition which could have resulted from closure of the component cooling surge tank vent valve on a high '

radiation signal. The installation of pipe and fittings in place of the relief valve establishes a low pressure drop overflow path to the waste disposal system. This change does not affect the operation of the component cooling system.

By removing the relief valve and installing piping, the barrier between the code class 3 component cooling piping and the non-safety related waste disposal piping is eliminated. This is considered acceptable per ANSI N18.2A-1975/ANS-51.8(4)(C). The material used for the spool piece

' is QA III schedule 40 carbon steel pipe which corresponds to the material of the waste disposal piping.

l i

l, 1

A-4

.. ~ . . . . . _ - . . . - . _ _

~ - - - . _ - - - - . - - - . - . . - . . - . - - -. . . . - - .- -