ML19294B478

Annual Rept of Occupational Exposure

ML19294B478
Person / Time
Site:	Monticello
Issue date:	02/28/1980
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML112990391	List: ML112990390 ML112990392 ML19294B478
References
NUDOCS 8003030401
Download: ML19294B478 (1)

Text

'

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

TOT" MAN-REM CONTRACT CONTRACT STATION UTILITY WORKERS AND STATION ILITY WORKERS AND WORK & JOB FUNCTION l

DIPLOYEES DIPLOYEES OTHERS EMPLOYEES u@LOYEES 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 & ENCR. 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 & CO?TEROLS 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.}39

• DESCRIPTION:

1.

Maintenance performed in Primary Containment during plant shutdown.

2.

Hanger & Support Modification 3.

Fuel Pool Modification 4.

Security System Installation tW63436VQL

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

f FDhTICELLO hUCLEAR 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 10CFR50.59(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.

Sumary 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.

EVALUATIN OF ENVIRONMENTAL QUALIFICATION OF SV-2790 (SRI 198)

Description of Change Review of the environmental qualification infomation 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 fomally qualified.

Sumary of Safety Evaluation The valve (ASCO #THF-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.

f 3.

REIDCATIm 0F SERVICE WATER RADIATION hDNI'IOR (76M071)

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 si6e stream sampling system was installed to allow this relocation.

Sumary of Safety Evaluation Relocation of this monitor does not alter its function.

In the event of side stream sampler failure, a low sample flow alam 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 TIE DESIGN LIFE OF TIE REACTOR VESSEL FEEDWATER N0ZZLES (77M063)

ADDENIXf! 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.

Sumary 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 perfomed 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 Sumer 1976 Addendum.

5.

NEW AININSTRATION BUILDING (782006)

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 requirenents 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.

e Summary of Safety Evaluation The new addition was designed and constructed in accordance with the latest revision of the Unifom Building Ccde 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 Unifom 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 perfomed 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 nomal source prior to the.r.odification. All modification work was completed in accordance with the original code requirements.

7.

FDDIFY HSMR SUPPORTS TO ACCEPT HSSA SUPPORTS (78M030)

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.

e 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.

RMIJ ISOLATION LOGIC hDDIFICATION (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.

Sumary of Safety Evaluation The tube wall thickness reduction is insignificant. The inner wall is not a structural manber. 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.