ML112990390

Forwards Annual Rept of Occupational Exposure, Rept of Changes to QA Plan & Annual Rept of Changes,Tests & Experiments

ML112990390
Person / Time
Site:	Monticello
Issue date:	02/28/1980
From:	Mayer L Northern States Power Co
To:	James Keppler NRC/OI/RGN-II/FO
Shared Package
ML112990391	List: ML112990390 ML112990392 ML19294B478
References
NUDOCS 8003030400
Download: ML112990390 (7)

Text

REGULATO INFURMAIiuN S TION TEM (- IDS)

DOC.DATE: 80/02/28 NUARILED: NO QOCKET #

ACCESSION NdR:8003030400 FACIL:50-263 Monticello Nuclear Generating Plant, Northern States 05000263 AUTHNAME AUTHOR AFFILIATION MAYER,O. Northern States Power Co, RECIP.NAME RECIPIENT AFFILIATION KEPPLERJG, Region 2, Atlanta, Office of the Director

SUBJECT:

Forwards "Annual Rept of Uccupational Exposure," rept of changes to QA plan & annual rept of changeSetests &

experiments.

DISTRIBUTION CODE: A008S COPIES RECEIVED:LTR j ENCL L S TITLE: Annual, Semi-Annual & Monthly Operating Reports (OL Stag NOTES: -- - - - - - - - - - - - - - - -

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME L.TrR ENCL ID CODE/NAME LTTR ENCL ACTION: 05 BC DOR6*3 6 6 INTERNAL: 0 1 1 02 NRC POR 1 2 14 "iPA 1 2 12 2 16 DIR OR 18 ENGR BR 1 1 1 19 REAC SAFT BR 1 20 PLANT SYS BR 1 1 1 21 EB 1 22 CORE PERF BR 1

23 EFFL TR SYS AEOD EX TERNAL: 03 LPDR 1 1 04 NSIC 1 1 24 NATL LAB 1 1 25 BROOKHAVEN 1 1 26 ACRS 15 15 MWAR 4 199 TUTAL NUMBER OF COPIES iiEQUIRED: LTTH 39 ENCL 39

0 0 NORTHERN STATES POWER COMPANY MINNEAPOLIS, MINNESOTA 55401 February 28, 1980 Mr J G Keppler, Director, Region III Office of Inspection & Enforcement U S Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137

Dear Mr Keppler:

MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Annual Report of Occupational Exposure and Changes, Tests & Experiments January 1 - December 31, 1979 Attached you will find two copies of the following reports:

1) Annual Report of Occupational Exposure

2) Annual Report of Changes, Tests, and Experiments These reports satisfy the annual reporting requirements contained in Section 6.7.A.2 of Appendix A to DPR-22 and Section 50.59(b) of 10CFR Part 50.

Yours very truly, L 0 Mayer, PE Manager of Nuclear Support Services LOM/DMM/jh cc: Director, IE, USNRC (c/o DSB) (40)

G Charnoff.

MPCA Attn: J W Ferman Attachment

II NUMBER OF PERSONNEL AND MAN-REM BY WORK AND JOB FUNCTION 1979 NUMBER OF PERSONNEL (1100 mrem) TOTAL MAN-REM .

CONTRACT CONTRACT STATION UTILITY WORKERS AND STATION UTILITY WORKERS AND WORK & JOB FUNCTION EMPLOYEES EMPLOYEES OTHERS EMPLOYEES EMPLOYEES OTHERS REACTOR OPERATIONS & SURVEILLANCE OPERATING PERSONNEL 38 0 0 31.804 0 0 HEALTH PHYSICS PERSONNEL 10 0 0 5.933 0 0 SUPERVISORY & ENGR. PERSONNEL 5 0 23 0.794 0 2.464 INSTRUMENT & CONTROLS PERSONNEL 7 0 0 3.414 - 0 0 SECURITY PERSONNEL 0 0 1 0 0 0.103 MAINTENANCE PERSONNEL 33 5 35 7.907 0.520 1.709 ROUTINE MAINTENANCE OPERATING PERSONNEL 16 0 3 2.028 0 1.615 HEALTH PHYSICS PERSONNEL 9 0 0 2.670 0 0 SUPERVISORY & ENGR. PERSONNEL 3 0 7 0.157 0 0.347 INSTRUMENT & CONTROLS PERSONNEL 6 0 0 0.734 0 0 MAINTENANCE PERSONNEL 29 7 21 22.395 0.576 5.439

• SPECIAL MAINTENANCE OPERATING PERSONNEL 2 0 0 0.165 0 0 HEALTH PHYSICS PERSONNEL 3 0 0 0.538 0 0 SUPERVISORY & ENGR. PERSONNEL 1 1 27 0.058 0.129 9.658 MAINTENANCE PERSONNEL 16 2 0 4.011 1.591 19.331 WASTE PROCESSING OPERATING PERSONNEL 8 0 5 1.419 0 4.589 HEALTH PHYSICS PERSONNEL 8 0 0 1.044 0 0 SUPERVISORY & ENGR. PERSONNEL 0 0 1 0 0 0.134 MAINTENANCE PERSONNEL 16 2 0 1.432 0.110 0 REFUELING OPERATING PERSONNEL 16 0 0 1.257 0 0 HEALTH PHYSICS PERSONNEL 3 0 0 0.115 0 0 SUPERVISORY & ENGR. PERSONNEL 1 0 0 0.052 0 0 INSTRUMENT & CONTROLS PERSONNEL 1 0 0 0.019 0 0 MAINTENANCE PERSONNEL 16 2 0 1.692 0.144 0

• • TOTAL OPERATING PERSONNEL 80 0 8 36.673 0 6.204 HEALTH PHYSICS PERSONNEL 33 0 0 10.300 0 0 SUPERVISORY & ENGR. PERSONNEL 10 1 31 1.061 0.129 12.603 INSTRUMENT & CONTROLS PERSONNEL 14 0 0 4.167 0 0 MAINTENANCE PERSONNEL 110 18 56 37.437 2.941 26.479 SECURITY PERSONNEL 0 0 1 0 0 0.103 GRAND TOTAL: 247 19 96 89.638 3.070 45.389

• DESCRIPTION: 1. Maintenance performed in Primary Containment during plant shutdown.

2. Hanger & Support Modification

3. Fuel Pool Modification 4. Security System Installation

• • INDIVIDUALS MAY BE LISTED UNDER MORE THAN ONE WORK AND JOB FUNCTION.

MONTICELLO NUCLEAR GENERATING PLANT ANNUAL REPORT OF CHANGES, TESTS, AND EXPERIMENTS 1979 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 10CFRSO.59(b).

1. ISOLATION OF THE CRD STABILIZING ASSEMBLY (SRI 197)

Description of Change The CRD stabilizing assembly was isolated to eliminate the carbon steel stabilizing assembly outlet piping as a contributor of corrosion products in the CRD cooling water line.

Summary of Safety Evaluation Testing conducted in April, 1978, verified that isolation of the stabilizing assembly has a negligible affect on normal CRD insert and withdraw. The CRD scram subsystem is completely independent of the stabilizing assembly, therefore, isolation of the assembly has no affect on the CRD scram function.

2. EVALUATION OF ENVIRONMENTAL QUALIFICATION OF SV-2790 (SRI 198)

Description of Change Review of the environmental qualification information for safety related electrical equipment located inside the containment revealed that the pilot solenoid valve (SV 2790) for the inboard reactor water sample isolation valve (CV 2790) was not formally qualified.

mSumary of Safety Evaluation The valve (ASCO #THT-8317A23) construction and operation were analyzed with the conclusion that the isolation valve would perform its function in the event of a LOCA. The solenoid valve is de-energized immediately upon detec tion of the accident (Group 1 isolation), vents the isolation valve diaphragm, and is not required to function thereafter. The isolation valve is closed and maintained in this position by spring force. Exposure to the LOCA environ ment is not expected to cause a mechanical failure of the solenoid that could prevent operation of the valve. Subsequent failure of the rubber seat materials (Buna-N) of the valve could result in air leakage from the pilot exhaust port but the design of the valve would prohibit re-pressurizing the isolation valve diaphragm. Thus the isolation valve would not reopen. This valve will be replaced with a fully qualified valve during the 1980 refueling outage.

3. RELOCATION OF SERVICE WATER RADIATION MONITOR (76MO71)

Description of Change The service water process radiation monitor was moved from a high background radiation location near the primary containment suppression chamber to the south-east corner of the Reactor Building at ground level elevation. A side stream sampling system was installed to allow this relocation.

Summary of Safety Evaluation Relocation of this monitor does not alter its function. In the event of side stream sampler failure, a low sample flow alarm is annunciated in the control room. Provisions have been made for obtaining a grab sample of the effluent in the event of side stream sampler failure. The plant service water is dis charged to the circulating water discharge canal which is continuously monitored, thereby, minimizing the probability of an unmonitored release.

4. EXTEND THE DESIGN LIFE OF THE REACTOR VESSEL FEEDWATER NOZZLES (771063)

ADDENDTIUM II Description of-Change The reactor vessel feedwater nozzles were reanalyzed to allow operation beyond the two year period analyzed in the revised feedwater nozzle stress report prepared by General Electric Company in 1977.

Summary of Safety Evaluation The revised stress report, prepared by NUTECH, shows that the feedwater nozzles have a design life of eight years of operation. A leakage monitoring sur veillance test is performed on a periodic basis to assure that actual nozzle fatigue usage is accounted for. The revised stress report satisfies the re quirements of ASME Boiler and Pressure Vessel Code,Section III, 1974 Edition with Addenda to and including the Summer 1976 Addendum.

5. NEW ADMINSTRATION BUILDING (78Z006)

Description of Change A new addition and remodeling of the existing plant administration building to accommodate an increase in plant staff and changes in administrative requirements was 95% completed. The 28' x 116' building addition is adjacent to the east side of the original administration building and has a full base ment and three stories.

Summary of Safety Evaluation The new addition was designed and constructed in accordance with the latest revision of the Uniform Building Code and the American Institute of Steel Construction specification for the design, fabrication and erection of structural steel for buildings. The new addition was designed for Uniform Building Code Zone I seismic acceleration which was the design criteria for the original building. To assure that the new addition does not interact with the original structure during an SSE event, a two inch separating joint was constructed between the two structures.

6. CRD PUMP SUCTION REROUTED TO CONDENSATE REJECT LINE (78Z027)

Description of Change A line was installed to provide high purity deaerated water to the CRD system from the condensate deminerlizer reject line. This modification will reduce the probability of CRD collet retainer tube cracking (tests performed by General Electric Company indicate that there is a direct relationship be tween dissolved oxygen concentration and time to failure of furnace sensitized stainless steel which is used in collet retainer tube construction).

Summary of Safety Evaluation This modification does not affect operability of the CRD hydraulic system. In the event the condensate system is removed from service the CRD pumps will automatically draw water from the Condensate Storage Tanks which was their normal source prior to the modification. All modification work was completed in accordance with the original code requirements.

7. MODIFY HSHR SUPPORTS TO ACCEPT HSSA SUPPORTS (78MO30)

Description of Change The support struts for certain HSMR hydraulic snubbers were modified to allow interchangeability with HSSA hydraulic snubbers.

Summary of Safety Evaluation Interchanging HSMR and HSSA hydraulic snubbers was verified to be acceptable by the Bergen-Paterson Pipesupport Corporation. The strut adapter pieces were designed and fabricated in accordance with Section III of the ASME Boiler and Pressure Vessel Code, Summer 1977 Addenda.

8. CRD SCRAM HEADER VENT CHECK VALVES (78M066)

Description of Change Check valves were installed in the CRD scram discharge volume vent lines to prevent backflow when a scram is reset.

-4 Summary of Safety Evaluation The check valves have no affect on normal operation of the CRD system or on operation during a scram. The valves were installed downstream of the vent isolation valves and are therefore not part of the pressure retaining boundary.

9. RWCJ ISOLATION LOGIC MDDIFICATION (79M058)

Description of Change The isolation logic for the Reactor Water Cleanup System was modified to initiate on high drywell pressure in addition to reactor low water level to provide diversity in the isolation parameters.

Summary of Safety Evaluation This modification did not degrade the existing isolation logic. It provides a more conservative approach to Reactor Water Cleanup System isolation as a single parameter is not relied upon to initiate the isolation function.

10. INSTALLATION OF HIGH DENSITY FUEL RACKS (79Z003)

Description of Change Eight additional High Density Fuel Storage System (HDFSS) modules were installed in the fuel storage pool per License Amendment No. 34 which was issued by the NRC on April 14, 1978. These modules differed from previous modules in the following three ways:

a) slightly reduced inner tube wall thickness.

b) central baseplate crossmember eliminated.

c) tube vent holes were left in corners of top and bottom.

Summary of Safety Evaluation The tube wall thickness reduction is insignificant. The inner wall is not a structural member. The effect on module Keff is minor and Keff will still be well below the 0.95 criterion. The baseplate cross members are not necessary load carrying members. They were not considered in the seismic analysis. The vent holes do not affect structural integrity. Corrosion test data indicate acceptable material loss over the projected 40 years life of the modules.