RA-98-008, Forwards 1997 Annual Insp Rept for Retired Westinghouse Test Reactor.Status of Retired Facility Remains Same as Time of Last Rept

Forwards 1997 Annual Insp Rept for Retired Westinghouse Test Reactor.Status of Retired Facility Remains Same as Time of Last Rept

ML20199B698
Person / Time
Site:	Waltz Mill
Issue date:	01/20/1998
From:	Nardi J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Michaels T NRC (Affiliation Not Assigned)
References
RA-98-008, RA-98-8, NUDOCS 9801280372
Download: ML20199B698 (7)

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y Westinghouse'ElectricCompany, Energy Systems som as6 aevieleno:ces corporanon Pittstugh PennM 15230 4 56 RA-98-008 January 20,1998 U. S. Nuclear Regulatory Comm!stion '

Office of Nuclear Reactor Regulation l

- Washington, DC 20555 -

Attention: Mr. Theodore S. Michaels

Reference:

Annual Repr & License

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- aet 50 22

Dear Mr. Michaels:

The Westinghousa Electric Company transmits herewith the attached " Annual Inspection of Retired WTR per License TR 2" for 1997. This report has been prepared in compliance with ~

the requirements of Facility License No. TR-2, Docket 50-22. The status of the retired facility -

remains the same as it was at the time of the last report.

If you have any questions regarding this matter, please write tr.3 at the above address or telephone me at (412) 374-4652.

' Sincerely,

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A.Joi.ep Na _ , Superv sory Engineer Regulatory Affairs -

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/wr Attachment Copies Transmitted: 3 notarized and 10 conformed

/h'dO COMMONWEALTH OF PENNSYLVANIA SS:

CGUN TY OF ALLEGHENY Swom and subsc.ribed before me this 22M day of hw4u/ ,1998.

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'1997 ANNUAL INSPECTION REPORT FOR RETIRED WESTINGHOUSE TEST REACTOR U.S.N.R.C. LICENSE TR 2 DOCKET 50-21 During the period September 29 thrcugh October 1 and on December 19,1997, the facilities of the retired Westinghouse Test Reactor (WTR), retained under NRC License TR 2, were entered for the purpose of conducting the annual survey, inspection, and for performing any preventive maintenance required. Those personnel entering the facilities to conduct the inspections were: R.G. Cline, Manager, Industrial Hygiene, Safety, and Environment .1pliance, R.L. Bussard, Manager, Facilities Services; 1

W.D. Vogel, Radiation Safety Officer; D.S DeArmit, Se;urity Officer, W.I. Rummel, Safety Engineer, J.T. DiNicola, Senior Radiological Techni. an; and J. Shouse, Senior Radiological Technician.

1 The findings anci actions taken are summarized below:

1 >

l) Dwing the past year, se /eral entries were made into the vapor containment for general inspections and to determine if significant wall condensati.v. or water was a ccumulating on the floor. Actions were previously taken to reduce humidity ar:d water accurautation including cove-ing the canal with plae. tic sheeting and polyurethane bonds; installing two (2) dehumidifiers - replaced in 1994: and installing a sump pump in a small hole (sump)in a low area of the floor. These actions continue to prove effective, and the humidity within the vapor container is being maintained at a very low level, and essentially no accumulation of water has been observed on the floor during the various inspections.

2) The general condition of the vapor container was good, and no significant moisture was observed on any interior walls, either above or below ground level. Visual inspection of the interior surface showed little or no increased deterioration of the surface coatin3. Some rusting was observed on the interior bottom portion of the vapor shell skin, but this condition was essentially unchanged from the last several years. Therefore. no action is deemed necessary concerr.ing ta:

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structural stability of the vapor container. The exterior surface of the vapor container was repainted in July 1992, and its overall condition remains generally good.

3) The water level in the canal was measured and found essentially the same as that observed during the pr-vious annual inspection (1996). The canal low level water alarm annunciator system was replaced in 1995. The canal low-level water alarm was manually activated, and the alarm responded properly, i.e., audible alarm and piintout in the Security Control Center and audible and visual alarm in the Health Physics office.

No work was done under the water in the vapor container canal since the last annual inspection so the conditio is remain unchanged. As previously reported , a system was installed in September 1986 to parmit the ion exchange processing of the canal water and return of the processed water to the car.al. The processing system continues to work well, and approximately 326,790 gallon. nas processed and recycled in 1997.

l As noted in Table 1 of this ryort, the canal water was sampled and found to have a gross beta-gamma activity of 4.4 X 10-7 pCi/ml. The canal water sample was analyzed for specific radionuclides and gav : me following results:

Nuclide M97 1996 1995 1994 4993 pCUml pCUml pCUml pCUml pCUml Cobalt-60 1.4 x 10-7 2.0 x 10-7 1.8 x 10 1.4 x 10-7 1.8 x 10-6 Cesium-137 4.7 x 10-7 4.1 x 10-7 1.6 x 10-7 8.4 x 10-7 6.9::10-7 Cesium-134 Not Detected Not Detsted Not Detected Not Detected Not Detected Stromium-90 3.3 x 10-8 1.4 x 10-7 5.2 x 10-8 3.6 x 10-7 2.8 x 10 '

4) Overall housekeeping within the vapor container remains satisfactory. All loose floor tiles have been removed and very little debris was observed on the floor.

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5) The personnel entry doots on the east and w est sides of the vapor container were found locked a the time of entiy, as were the doors leading into the Rabbit Pump room, the Sub-Pile room, and the cover over the primary coolant pipe tunnel. These areas were unlocked to permit access for this inspection but were again locked when the inspection was completed.

6) , The valve on the drainline on the bottom of the reactor (inside the Sub-Pile voom) was inspected.

While the valve was removed during the 1993 characterization to allow access to the interior of the vessel, the remaining piping was verified to be open to the interior of the vessel thus assuring venting of the vessel. A channel was installed to divert condensation from within the vessel to the canal. The butterfly valve in the ventilation ducts in the Truck Lock Area was inspected and found in the CLOSED position.

, 7) The absolute filter through which the reactor primary coolant system breathes into the Annex area was inspected and found in good condition. The filter was replaced in March 1996.

8) As previously noted, several entries were made into the vapor container during the year. The key for the area is maintained by the Site Security Guards, and records of all entries, indicating date and time, purpose and names are maintained by the Radiation Safey Officer.

9) The following changes were made in Westinghouse management relative to the administrrution of License TR 2 since the submission of the 1996 aanual report.

Mr, Bruce Bevilacqua has replaced Mr. J.J. Bastin as the Manager Resources and S.ippea Operations. Mr. Bevilacqua is the senior site manager having responsibility for the retired WTR facilities.

R.G. Cline remains as the Manager, Industrial Hygiene, Safety, and Environmental Compliance (IHSEC). Mr. Cline reports directly to Mr. Bevilacqua.

W.D. Vogel, Radiati'n Safety Officer, reports directly to Mr. Cline.

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10) The inspection of the primary coolant tunnel (snakepit) surrounding the vapor container indicated that approximately 14,400 gallons of water, with a gross beta activity of 2.2 x 10-6 pCi/ml. The volume of water in the primary coolant tunnel remains essentially unchanged from 1996.

11) The results of the radiological survey are shown in Table 1, r.ttached. A review of the survey data obtained since 1963 indicate that the accesc% radiation and contamination levels are very low and relatively stable. Frequently, some scatter is seen in the data that is attributed to sampimg techniques, locations, sensitivity of the cc.unting instrumentation, radioactive decay, and relatively low levels of activity present. Any slight changes in a specific set of data from year-to-year are not considered sig'nificant.

12) U. ., c ssible locations inside the reactor vessel, and hardware in the canal have higher radiation levels, principally from metal activation of reactor components and surface contamination, i

I The sampling techniques, equipment, and instrumentation used during this inspection were as follows:

a) Water samples were taken in clean polyethylene bottles at the inlet to the canal ion exchange processing system or as " grab" samples in non-flowing systems. An aliquot of the sample was evaporated to dryness in a two-inch stainless steel planchet and counted in an automatic, thin.

window gas proportional, alpha-beta anti-coincidence counter. The counting efficiency for beta-gamma activity was 44.7 % and the background was 1.83 cpm.

For specific radionuclide determinations, an aliquot of the untreated sample was analyzed by gamma-ray spectrometry using a high resolution germanium detector. The strontium-90 activity was determined by performing a chemica! separation on a portion of the sample and then counting the separated strontium-90 fraction in the counter described above.

b) Contamination survevs were performed by taking random smears of approximately 100 cm2 using 3 cm filter paper (Whatman 5 or equivalent). The filter paper was then counted in the gas 98001/980105

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prohrtional counter described in (a), above. A varying number of smears were taken for each '

area listed in Table 1, and the average value was reported.

c) Air samoles were collected using a Staplex high volume sampler with the particles impkging on a !ightly greased stainless steel planchet. The sampler flow rate was approximately 40 cfm, and 4 the sampling time was 10 minutes or longer. After sampling, the planchet was counted in the gas proportional counter described in (a), above,

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d) Radiation surveys were made at random locations within each area shown in Table 1.

.. Measurements were made at waist height using an Eberline Model E-520 with a G M detector with a 30 mg/cm2 window, and a Ludlum Model 19 Micro-R meter with a Na! crystal.- The average radiation level for each area was deterrded and reported in Table 1.

L 48C31/980105

lABLEI

Reference:

WTR-172, Appendix A RADIATION AND CONTAMINATION AREAS SURVEY REPORT Retired WTR Facility License TR-2 Survey Dates: September 29 -October 1,1997 Health Physics Procedure No. 29 p-y Contamination Levels

• LOCATIONS Levels Surface Air Water i

mr/hr dpm/100 cm2 pCi/mi- pCi/ml j l. Reactor Building 1, 16 ft Elevation - <1 < 200 2.1 x 10-12 N.A.

General Background

2. Rabbit Pump Room - 2.0 < 200 1.1 x 1012 ' N.A.

General Background

3. Sub-Pile Room - 1.5 598 ' 9.1 x 10-12 N.A.

General Background

4. Reactor Top - <1 < 200 1.8 x 10-12 N.A.

- General Background

5. Reactor Top - .< 1 < 200 7.4 x 10-13 N.A.

Over Closed Vessel-General Background

6. Reactor Top - <1 < 200 . 8 x 10-12 N.A.

Around Trench -

General Background

7. Canal Wall Top <l N.A. N.A. N.A.

8. Canal Water <1 N.A. N.A. 4.4 x 10-7

9. Pit - <1 < 200 3.6 x 10-13 2.2 x 10-6 PC Pipe -

Tunnel 16 ft Elevation

• Average values N.A.' Not Applicable

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