IR 05000344/1997003
| ML20210S731 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 09/09/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20210S698 | List: |
| References | |
| 50-344-97-03, 50-344-97-3, 72-0017-97-03, 72-17-97-3, NUDOCS 9709120047 | |
| Download: ML20210S731 (18) | |
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s ENGJASDEE_2 U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket No.-
50-344; 72 17 License No.:
NPF-1 Report No.:
50 344/97-03; 72-17/97-03 Licensee:
Portland General Electric Company (PGE)
Facility:
Trojan Nuclear Plant
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Location:
121 S. W. Salmon Stroe;, TB-17 Portland, Oregon
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Dates:
July 21-24,1997 I
Inspectors:
J. V. Everett, Health Physics Inspector
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M. Haque, Project Manager Accompanied By:
D. Blair Spitzberg, Ph.D., Chief Nuclear Materials Inspection and Fuel Cycle / Decommissioning Approved By:
D. Blair Spitzberg, Ph.D., Chief Nuclear Materials Inspection and Fuel Cycle / Decommissioning Attachment:
Supplemental Information 4709120047 970909
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PDR ADOCK 05000344 G
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2-I EXECUTIVE SUMMARY Trojan Nuclear Plant NRC Inspection Report 50-344/97-03 The Trojan nuclear facility was shutdown in November 1992. The licensee had developed a decommissioning plan and initiated dismantlement and decontamination activities at the facility. Trojan had completed significant work in removing contaminated material from the f acility and shipping the material to Hanford,- in central Washir.gton, for burial. The major components in the containment had been removed with the exception of the reactor
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vessel. The proposrl by PGE to ship the reactor vessel to Hanford as a si-lo unit was being reviewed uy the NRC. Completion of the decommissioning of the facuity was estimated to be the end of 2001.
In addition to the effort to decontaminate and dismantie the reactor f acility, PGE had applied for a Part 72 license to place the spent iuel currently in the spent fuel pool into dry cask storage at an onsite Independent Spent Fuel Storage installstion (ISFSI). Trojan planned to use a cask provided by the Sierra Nuclear Corporation that will be licensed for both storage and transport. The NRC licensing review process of the cask was underway.
Movement of spent fuelinto the ISFSI was scheduled for completion by late 1998 or early 1999.
Control of Radioactive Material e
The licensee had established a work area, procedural process, and release limits for the free release of material from the radiologically controlled area. The free release program was reviewed with consideration of a recent licensee identified incident in wnich slightly contaminated material had bet.n released from the radiologically corarolled area. One violation was identified involving f ailure to perform adequate surveys of potentially contaminated materials as required by 10 CFR 20.1501 (Section 1).
Soent Fuel Pool Debris Project Work sorting fuel debris in the spent fuel pool was underway. Health physics
controls were in place, and health physics coverage was provided for any actions involving removal of material from the pool (Section 2).
Drv Cask Storaae
The installation of reinforcement steel and placement of cor crate for the ISFSI pad had been performed in accordance with site specifications. The ISFSI pad design had appropriately considered the potential for soilliquefaction through soit review and analysis of soil boring samples (Section 3).
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3-Spent FueLP_os]
Spent fuel pool water level and temperature had been maintained within the required
Technical Specification for the period of April 1 through July 15,1997 (Section 4),
Bodioloaical Environmental _ Monitorjng The 1996 Annual Radiological Environmental Monitoring Report and the 1996
Annual Radioactive Effluent Release Report were reviewed. No unusual or abnormal radiological conditions were identified during 1996 for the Trojan site.
Environmentallevels were consistent with levels reported over the past 5 years (Section 5).
Health Physics Activities Health physics controls for both radiological exposures and for contamination were
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similar to exposures in past years (Section 6).
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Emolovee Concern Proaram The licensee's employee concern program appeared to ba receptive to employees o
raising safety and quality issues. Comments by employees suggested that PGE management was receptive to employee identified safety and quality problems.
First line supervision had been the main avenue for resolution of day-to-day problems. This appeared to have been an effective process. However, some weaknesses were noted in the licensee employee concern program related to feedb ck provided to the workers concerning final resolution of issues. This was discussed with the licensee (Section 7).
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Ilea nt d etaih Summarv of Plant Status A significant amount of dismantlement work had been completed at Trojan since the last inspection. Work inside containment was progressing with removal of internal concrete walls, structural supports, and preparation for removal of the reactor coolant system main loop piping inside the bioshield. The opening in containment that had been used to remove the steam generators was being enlarged to facilitate removal of the reactor vessel.
Dismantlement work in the auxiliary building anti fuel building included the boric acid injection tank, service water booster pumps, and radwaste evaporator. Removal of the residual heat rernoval heat exchangers (both A & B) was nearing completion. A new free release survey area had been designated since the last inspection. The area appeared to be well organized for free release activities.
The effort to segregate the fuel debris filters in the spent fuel pool was underway. During the removal of a filter from the spent fuel pool, a fuel pellet was unknowingly brought to the surface. Af ter removal of the fuel pellet from the filter, the pellet was returned to the pool. Health physics controls were effectively implemented during the process resulting in only 1 mR exposure for each of the two individuals who sorted and returned the fuel pellet to the fool.
The storage pad for the ISFSI was under construction. The pad consisted of two parts:
one area for storage of the casks and a second area for cask transfer activities. Concrete for halt of the storage pad had been poured the week prior to this inspection. The second half was poured during this inspection. The rebar placement and the design of the pad were reviewed and were found to be consistent with site specifications.
Control of Radioactive Material (83726)
1.1 Insoection Scoce Contaminated material was being removed from the Trojan f acility for either burial as radioactive waste or decontaminated and surveyed for free release from the site.
The free release program helped reduce the radioactive waste volume and costs associated with dismantlement. The process and release criteria used for the free release program were reviewed to deteimine if adequate controls had been i
established to prevent release of radioactive material.
1.2 Observations and Findinas Trojan had an aggressive program to free release material from the site if decontamination could be accomplished in a cost effective manner. This effort reduced the amount of material requiring burial as radioactive waste, which in turn reduced costs to PGE. Much of the released material was sold as scrap. During dismantlement, material was sorted based on whether contamination was present or-
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i the item was relatively uncontaminated. Material that was initially determined to be uncontaminated was sent to a free release area und surveyed in a low background -
area prior to placement in a scrap _ metal bin Material that was suspected or known to be contaminated was sent to a decontamination area to either be decontarninated for free release or placed in radioactive waste containers for shipment to Hanford for burial. Of the material sent directly to the free release area, approximately 10 to-15 percent was typically found to have smalllevels of contamination still present, which required cleaninn prior to free release. The material found acceptable for free release was placed in a scrap metal bin outside the radiologically controlled area.
The scrap metal bin was surveyed with a microR/hr meter prior to removal from the Trojan sito. This provided a secondary check. This survey was capable of finding gamma radiation levels above background, but would not be effective in detecting very low levels of contamination that could be on the scrap metal.
Free release surveys were conducted using Trojan Procedure TPP 2019, " Release of Materials From Radiologically Controlled Areas," Revision 0, dated August 30, 1993. Section 6.1 of this procedure established the free release criteria of 1,000 disintegrations / minute (dpm) per 100 cm loose beta or gamma contamination,
2 1,000 dpm/20 cm fixed beta or gamma contamination, and no alpha contamination.
The limits for beta and gamma radiation were equivalent to the limits established in NRC Regulatory Guide 1.86, " Termination of Operating Licenses for Nuclear Reactors." The licensee's limits for alpha contamination were more conservative than the limits in Regulatory Guide 1.86. Trojan Procedure TPP 2019 also required the released material from the radiologically controlled area to be documented on RP Form 205, " Free Release log."
The Nuclear Enterprise gas proportional counter probe model No, CM 11 was used for free release surveys. The CM 11 is a large area probe of 100 cm' which can be detached from the base unit to allow for better mobility. The probe could be used for approximately 15 minutes before it had to be retumed to the base unit to be recharged with gas. The probe had a low count rate alarm setting which would provide for an alarm when the probe was not detecting normal background radiation levels and, therefore, needed to be recharged with gas. Trojan Procedure RP 94,
" Contamination Monitor CM 11," Revision 1, dated September 23,1996, provided instructions for using the CM 11 probes. This procedure established a survey frisking speed of 2-3 inches per second with the probe held no further away than 1 inch from the item being surveyed. The procedure also specified that af ter approximately 15 minutes of continuous monitonng, the probe needed to be returned to the base unit to be recharged. An attachment to the procedure provided calibration settings for the CM 11. This attachment specified a zero setting for the low count rate alarm.
During inspection 97-02 conducted April 14-17, 1997, the Trojan health physics supervisor informed the NRC inspector onsite that a senior radiation protection technician had been reported by co-workers as not following the procedural requirements for surveying material _being free released from the radiologically
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6-controlled area. The technician had been observed performing the required frisk at a speed greater than specified in procedure RP 94. A subsequent check of the material released from the radiologically controlled area found several pieces of scrap metal that exceeded the free release limits of procedure TPP 20-19. In addition, not all the materialin the bin had been properly logged on a free release form. Trojan took immediate actions in response to the event by removing the individual from free release survey responsibilities and initiated a comprehensive investigation into the circumstances surrounding the release of the contaminated
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material from the radiologically controlled area. The licensee's review of the
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incident was thorough and complete. This non-repetitive, licensee identified and corrected violation is being treated as a Non-Cited Violation, consirient with Section Vll.B.1 of the NRC Enforcement Policy, l
Additionalinvestigation into the incident by the radiation protection manager found that one of the CM 11 probes used at the free release area did not function reliably l
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when tested by the radiation protection manager. Trojan initiated an internal i
investigation and opened Corrective Action Request CAR 97-06. During an i
evaluation of the operability of the CM 11 probes, Trojan discovered that by setting the CM-11 probe's low count rate alarm to zero, per Procedure RP 94, that the alarm was effectively disabled for situations where the gas was low in the instrument and background was no longer being detected. The proper setting should have been one count /second. The CM-11 probes were taken out of service until the radiation protection group determined whether to continue use of the instruments. The licensee determined that problems with the CM 11 probe may have contributed to the release of several slightly contaminated pieces of scrap metal from the radiologically controlled area. The contaminated material was subsequently decontaminated and released as scrap. Trojan initiated an intemal investigation to determine if other contaminated material may have been released.
Surveys are required by 10 CFR 20.1501(a)(2) which states, in part, that each licensee shall make or cause to be made, surveys that are reasonable under the circumstances to evaluate (i) the extent of radiation levels; (ii) concentrations or quantMies of radioactive material; and (iii) the potential radiological hazards that could be present. Based on the licensee's investigation findings relating to use of the CM 11 probes, the NRC has determined that a violation of 10 CFR 20.1501(a)(2) occurred involving a f ailure to perform adequate surveys of material being released from the radiologically controlled area. This violation will be tracked as Violation 50 344/97003 01.
1.3 Conclusion The licensee had established a work area, procedural process, and release limits for the free release of material from the radiologically controlled area. The free release program was reviewed with consideration of a recent licensee identified incident in which slightly contaminated material had been released from the radiologically
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7-cantrolled aea. One violation was identified involving f ailure to perform adequate surveys of potentially contaminated materials as required by 10 CFR 20.1501.
Spent Fuel Pool Debris Project (84101)
2.1 Inso. cction Scoce The licensee had initiated a project to sort and solidify debris material, filters, and fuel pellets that were stored in the spent fuel pool to allow transfer of the material to dry storage in the ISFSI. This material will be placed in cans and heated with superheated steam to 1100*F to drive off organic material that could decompose over time and form hydrogen in the sealed can. The health physics controls and radiation work permit provisions for the planned activities were reviewed.
2.2 Observation and Findinas
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The spent fuel pool contained not only spent fuel assemblies, but also a number of other components and debris that had been placed in canisters in the pool for a
shielding and housekeeping purposes. The canisters in the pool contained nonfuel
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bearing components, such as parts of spent fuel assemblies that did not contain I
fuel, fuel debris which included fuel pellets, fragments, or portions of spent fuel assemblies that could contain spent fuel, and debris _which included miscellaneous equipment, material, and filters in addition, material had f allen to the bottom of the spent fuel pool over the years. The licensee removed a section of the spent fuel pool rack and vacuumed the bottom of the pool. Saven fuel pellets were collected plus a number of fuel pellet pieces.
During the inspection, nonspent fuel assembly material was being segregated between nonfuel bearing components, low level waste, and potential gas generating materials. The material was being sorted on a table near the bottom of the fuel transfer canal, approximately 20 feet underwater. An underwater camera was used
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to assist in determining the identity of the material being sorted. The material that
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was sorted for processing in the steam reformer was placed in canisters and stored
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in the fuel transfer canal. Material designated for low level waste disposal was vacuumed into an underwater vacuum cleaner. Periodically, the vacuum cleaner filter was brought to the surface and the contents emptied into a bag for low level waste.
Radiation work permit RWP 97-14, Spent Fuel Pool Debris Cleanup, Revision 1, dated May 28,1997, and the associated radiation protection technician work
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instruction was reviewed. The radiation work permit established monitoring requirements, contamination levels in which decontamination would be required, dose rate limits in the work area, and limits for material which can be processed.
The process canisters were limited to 350 R/hr. Material placed in the low-level waste bag was limited to 200 mR/hr. Protective clothing and finger rings were
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required for the work involving sorting of the material. The radiation work permit
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~8-required that a radiation protection technician be present with a survey instrument whenever material was removed from the pool.
NRC Information Notice 90 33, " Sources of Unexpected Occupational Radiation Exposures at Spent Fuel Storage Pools," issued May 9,1990, was reviewed with the radiation p itection manager. Problems described in the information notice were applicable to we. underway at Trojan to sort the radioactive material underwater.
Controls recommended in the information notice had been evaluated previously by PGE and applicable proWsions incorporated into procedmas. The primary control used by Trojan was the requirement for a radiation protection technician to be present any time material was removed from the spent fuel pool.
The estimated dose assigned to RWP 9714 was 1643 mrem. As of July 23,1997, 1815 mrem had been recorded against the radiation work permit. This was based on over 7000 hours0.081 days <br />1.944 hours <br />0.0116 weeks <br />0.00266 months <br /> worked.
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On July 23,1997, while bringing the vacuum cleaner basket to the surface of the spent fuel pool to empty the contents into the low level waste bag, the radiation protection technician detected higher than expected radiation levels. The readings exceeded the 200 mR/hr limit for placement of the materialinta the low-level radioactive waste bag. A reading of 800 mR/hr beta-gamma with the instrument beta shield open was measured on the vacuum cleaner basket as it was brought out of the water, The contents were emptied onto a table, and the material was sifted through using long handle tongs. The highest reading measured during the sorting process was 1 R/hr, open window, at 1 foot. At 7 fect, the reading was 1 mR/hr, indicating a high beta factor was involved. A complete fuel pellet and a partial fuel pellet were identified and placed back into the basket and returned to the pool. The digital alarming dosimeters worn by the workers measured 1 mrem. After the fuel pellet was returned to the spent fuel pool, work was stopped, and a meeting held to discuss the event, it was determined that in future situations where the dose rates exceeded 150 mR/hr on contact, the material should be returned to the underwater table for further sorting. Revision C of Work Order 11351 was issued with this change to Sections 4.5.1 and 4.5.5. In addition, a new work table was obtained which would be approximately 6 feet underwater to provide better viewing for conducting sorting activities.
2.3 Conclusion Work sorting fuel debris in the spent fuel pool was underway. Health physics -
controls were in place, and health physics coverage was provided for any actions involving removal of material from the poo.
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3 Dry Cask Storage (60851,60854,46051,40053)
3.1 Insoection_ Scope The licensee had progressed with their of' ort to establish an ISFSI located on the northeast corner of the current industrial security area. This inspection included observation of the pouring of concrete for a portion of the storage pad and review of the schedule for the remaining activities associated with the dry cask storage project.
3.2 Observations and Findinos The licensee planned to move the spent fuel from the spent fuel poolinto dry cask storage in the on site ISFSI starting ir, mid 1998. The work activities underway were related to the pouring of the concrete pad for the ISFSt. The pad consisted of two parts. One portion of the pad will be used for storage of the casks. A second portion of the pad will be used for a cask transfer station. The transfer station will be used if a leaking cask is detected and the cask needs to be moved into an overpack. The storage portion and the transfer portion of the pad had different design criteria.
Concrete had been poured for half of the storage pad area. Forms and placement of reinforcement steel had been completed for the other half of the storage pad.
Concrete was poured for that area during this inspection. The construction drawings were reviewed to verify the proper placement of the reinforcement steelin accordance with the drawings. Placement was determined to be consistent with the drawings.
A review of the " Structural Design of Concrete Storage Pads for Spent Fuel Casks,"
prepared by Anatech Research Corporation, was completed to verify that the site had been properly characterized as seguired by the provisions of 10 CFR 72.212(b).
The review included the licensee's documentation to ensure that mechanisms that could cause ISFSI pad failure had been considered and that the consequences of such events would not place the ISFSI, or its components, in an unanalyzed condition. In particular, reviews and analysis of soil boring were evaluated relative to the potential for liquef action of the pad site. The review and conclusions by the licensee appeared to be appropriate.
The licensee's quality assurance staff was actively monitoring construction activities. Specific procedures for pouring the ISFSI pad concrete and other related construction drawings were available at the site. During the review of the design drawings, construction specifications, and procedures for the work completed on the storage portion of the p:sd, the following observations were made:
Reinforcement steel was the correct size and grade, and correctly installed.
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10-Compressive strength samples of concrete were taken. Subsequent strength e.
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Concrete was properly placed in batches and consolidated, Overall dimensions, orientation, and the levelness of the pad were correct.
e The review of the pad construction activities found no inconsistencies with pad design and no problem with the construction of the pad, in addition to reviewing the pad construction activities, the schedule of other activitit-for the ISFSI were discussed with the licensee. The following table provides a listing of several key activities and estimated dates.
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l Fuel Inspection Complete Crane Test (Initial load test)
Complete Pad Construction Complete (Storage pad portion)
Complete Fue: Debris Project (System operational & tested)
Aug.18,1997 Storage Cask (Rebar in place & ready to pour concrete)
4th Q 1997 Transfer Cask (Start of Fabrication)
4th Q 1997 Basket (Start of f abrication)
4th Q 1997 Security System (Start of final operational testing)
1st Q 1998 Cask Welding Dry Run (Demonstrate welding of lids)
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1st Q 1998 Loading of First Cask 2nd Q 1998 Dates for the activities listed in the table are estimates and will be updated as supporting tasks are completed and as NRC design approvals are issued.
3,3 Conclusions The installation of reinforcement steel and placement of concrete for the ISFSI pad had been performed in accordance with site specifications. The ISFSI pad design had appropriately considered the potential for soil liquef action through soil review and analysis of soil boring samples
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Spent Fuel Pool (86700)
4.1 Insoection ScoR2 Technical Specification 3.1 established the requirements for the spent fuel pool water level and temperature. This inspection included a review of the spent fuel pool's water level and temperature records for the second quarter of 1997.
4.2 Qhsgrvations and Findinas Technical Specification 3.1.1 required the water level of the spent fuel pool to be maintained a 23 feet over the top of the irradiated fuel assemblies. Technical Specification 3.1.3 required the paol to be maintained at a temperature of s 140*F.
The records from April 1 through July 15,1997, were reviewed. The water level was maintained in excess of 24 feet over the top of the fuel assemblies throughout this period. The spent fuel pool temperature was maintained below 80*F.
4.3 Conclusion Spent fuel pool water level and temperature were maintained within the required Technical Specification for the period of April 1 through July 15,1997.
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Radiological Environmental Monitoring (80721)
5.1 Insoection Scooe Submittal to the NRC of annual reports for the radiological environmental monitoring program and the radiological effluent release program were required bv Technical Specifications 5.8.1.2 and 5.8.1.3. Trojan submitted both required reports to the NRC. This inspection included a review of the reports.
5.2 Observations and Fir dings a.
Annual Radioloaical Environmental Monitorina Reoort The Annual Radiological Environmental Monitoring Report for 1996 was submitted to the NRC on April 17,1997. The report included a summary of the radiological data collected for 1996 around the Trojan site. Sumpling included water, air, soil and ambient radiation levels. in addition, the results of the 1996 Environmental Protection Agency and Department of Energy interlaboratory comparison program results for Trojan data and PGE quality control analysis program resuits were provided.
Air narticulate samples were collected weekly and analyzed for gross beta activity
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from two on-site locations. One location was at the north site boundary. The other f\\
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location was south of the plant at the meteorological tower. Sample analysis j
indicated airborne levels consistent with levels detected over the past 5 years.
Well water samples were collected quarteriy from the Prescott Oregon, water
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supply. No elevated tritium or gamma omitting isotope levels were detected.
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Drinking water samples were collected at St. Helens, Oregon, which is upriver from
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Trojan, and at Rainier, Oregon, which is downriver. The drinking water sampler is a
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ccmpositing sampler that takes a sampic every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Every 4 weeks the water is i
collected for analysis of tritium, gross beta, and gamma emitters. The data collected for 1990 indicated normal background levels for gross beta with the tritium and gamma cinitters below the lower level of detection.
Ambient radiation levels were determined by 12 thermoluminescent dosimeters j
(TLDs) placed at vr. as locations around the site and at St. Helens and Rainier.
The TLDs were a"
, ad quarterly by an independent company. All dosimeters indicated normt,I L M Aground levels except for dosimeter No.15, which recorded an average of 0.19 m9/ day. This is approximately twice background. Trojan evaluated
j the high reading and determined that the dos! meter had measured radiation levels I
from the refueling water strage tank and from radioactive material stored in outside areas of the restricted area. Dosimeter No.15 was located on the Industrial area
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I Shoreline soil samples were collected from a location on the bank of the Columbia River near the Trojan site. Analysis was performed for gamma emitters. The
samples consisted of approximately 1 quart of soil taken over a 1 square foot area
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to a depth of 1 to 4 inches. Vegetation and large rocks were removed. Sampling was conducted twice a year. Both samples had no gamma emitters above the lower
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level of detection of 0.18 pCilgram Cs 137.
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AnnuaLEffluent Release Reoort
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The Annual Radioactive Effluent Release Report for 1996 was submitted to the NRC on April 21,1997. The report included a summary of the liquid, gaseous, and solid releases from the site. For 1996, no releases of airborno radioactivity occurred above the limits of 10 CFR Part 20, Appendix B, Table 2, Column 2.
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For liquia rueases, there were 53 batch discharges during 1996. The highest level of liquid radioactivity released during any one quarter, excluding tritium, was
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0.0379 Curie. This is well below the 2.5 Curio limit specified in section 3.2.1.2 of the offsite dose calculation manual. For tiitium, the highest release for any one quarter was 2.9 Curies. This represented less than 0.1 percent of the etfluent
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concentration value allowed for tritium from 10 CFR Part 20.
During 1990,44 shipments of solid radioactive waste were completed. Ali shipments were made to U. S. Ecology at Hanford, Washington The predominant
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13-isotopes in the shipments were Co 60, Cs 137, Fe 55, and Co 57. A total of approximately 350 Curies were shipped for disposal, 5.3 C2nclusiori The 1996 Annual Radiological Environmental Monitoring Report and the 1996 Annual Radioactive Effluent Release Report were reviewed. No unusual or abnormal radiological ef fluents were identified during 1996 for the Trojan site. Environmental levels were consistent with levels reported over the past 5 years.
6-Health Physics Activities (83100)
6.1 Insoection Scons Health physics activities associated with the decontamination and dismantlement work underway at Trojan was reviewed. This included observation of work in progress, review of exposure records for the first half of 1997, and review of the coritamination incidents to date.
6.2 Observations and Findinan Observation of health physics personnel performing activities in containment indicated that coverage was offective in ensuring that health physics requirements were being implemented. Conversations with the health physics supervisor, the radiation protection manager, and a health physics technician indicated that compliance with radiation work permit requirements by contract workers had been good. No significant problems had been noted during 1997. A tour of containment, the fuel building, and the auxiliary building indicated that areas were posted and barricaded as required, protective clothing was readily available near contaminated areas, and waste receptacles for contaminated clothing were positioned at step off pads. Radiation levels throughout the plant were posted on floor plans at the entrance to the radiologically controlled area. The data appeared to be current.
Several high radiation areas were identified on these floor plans and wem verified as properly posted and in compliance with Technical Specification 5.10.1 by touring the areas. Technical Specification 5.10.1 establishes posting, barricading, and locking requ4ements for high radiation areas.
Radiation levels throughout the plant continue to be low and were being reduced further as equipment was removed. Radiation levels in the auxiliary building wore typically below 1 mR/hr. General area radiation levels in containment were typically below 10 mR/hr. Contamination controls continue to be effective in minimizing the spread of contamination to new areas.
Eighteen radiation work permits were active. RWP 9712 was established for work associated with containment component removal / disposal. Accumulated dose as of July 23,1997, was 6575 mrem for 10,816 hours0.00944 days <br />0.227 hours <br />0.00135 weeks <br />3.10488e-4 months <br /> worked. This is less than
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- 14-1 mrern/hr average. This average is representative of most work underway at Trojan. The only job in which personnel radiation exposure averages exceeded 1 mR/hr was RWP 9716, " Reactor Vessel Preparation." This work involved being close to the reactor vessel in preparation for removing the piping from the vessel.
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Even though the reactor vecselis filled with water, the 2 million curies of estimated activity associated with activation of the vesselinternals created high radiation areas where ths work is being performed.
Exposure records for the first half of 1997 were reviewed. As of the end of June, the total dose on site was 17.6 man tem. There were 204 personnel actively
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l badged at Trojan. The highest dose recorded to date in 1997 was 528 mrem as of l
July 23,1997, by a contract craftsman. There were 49 individuals with doses over d
100 mrom. The doses were consistent with levels recorded in previous years. For 1996,59 workers exceeded 100 mrem with the highest dose at 1970 mrom. Total site wide exposure for 1996 was 44 man rem. This compares to 41.5 man rem exposure for 1995 of which 93 workers exceeded 100 mrem for the year.
Twenty.one contamination.'ncidents had been recorded for 1997. Of these, two were determined to be significant enough to justify further investigation and assignment of dose to the individuals. Both involved hot particles of Co 60 which had migrated through the protective clothing during sweating and had come into
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contact with the individuals' skin. The contamination was discovered by the
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personnel contamination monitors when exiting the area. The first person was involved with removal of scaffolding around the boric acid accur 'ulator tank. His
contamination was located at waste level. A germanium lithium detector count of the particle indicated 0.253 microcuries (uCi) of Co 60. Radiation levels from the particle were measured at 0.15 mR/hr gamma and 27.5 mrad /hr beta on contact.
The computer code VARSKIN was used to determine the dose. Based on a maximum stay time of 220 minutet the skin dose assigned to the individual was 3.61 rads.
The second individual was stringing lights in an area where contaminated pipe cutting activities were being conducted. His contamination was found on the upper thigh and was determined to be 0.121 uCi of Co 60. His assigned skin dose based on an exposure time of 180 minutes was 1.41 rad.
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6.3 Conclusion Health physics controls for both radiological exposures and for contamination were
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effectively being implemented. Personnel exposures for the first half or 1997 were
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similar to exposures in past years.
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Employee Concem Program (71707)
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7.1 Insoection Scops The employee concern program is a process in which employees can raise safety concerns to management for resolution. At Trojan, this program is called the Excellence Response Program. A review of the program was performed to determine if adequate provisions for making employees aware of the opportunity to raise sefety concerns existed.
i 7.2 Observation and Findinos
The Trojan Excellence Response Program has been an ongoing program to i
encourage employees to raise safety and quality concerns to management and to
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understand the options available to further report these concerns to the NRC should the employee feel that resolution of the nroblem by Trojan was inadequata. The program applied to both employees and contractors. All new personnel assigned to the Trojan site received training that included employee rights and the employee concerns program. The program was reinforced through annual training, in addition, postings were located throughout the site f acilities p'oviding phone numbers for the Excellence Response Program as well as the NRC.
The Trojan site executive and plant manager issued a memorandum to site employees in October 1990 discussing the NRC policy statement on raising safety concerns, in November / December of 1996, Trojan distributed a questionnaire to 29 selected individuals representing a range of plant organizations and responsibilities.
The questionnaire covered a nurrbor of questions concerning the quality assurance program. In addition, the questionnaire included two questions concerning awareness of the Excellent Response Program and the opportunity to raise quality concerns to management. All respondents were aware of the programs and felt they could raise concerns to management. There was some concern expressed in response to the questionnaire as to how site contract employees may view the Trojan employee concern program and whether they felt reluctant raising concerns.
During this inspection and several previous inspections, conversations were held by the inspector with several PGE employees and contract personnel while observing work activities and during breaks. There appeared to be a willingness by both the employees and the contract personnel to express their opinion to their first line supervisors concerning quality and safety issues, in no case was a reluctance to identify safety concerns found with the small group sampled. However, there appeared to be weak communications back to the employees as to actions taken in response to concerns expressed that required action beyond the worker's immediate supervisor. A recent problem with smoke in the containment was e good example.
Though PGE had initiated an assessment of the problem, there was little feedback to the contract personnel who expressed the concern as to what actions were being taken by Trojan to assess the problem and what the final determination by Trojan
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-16-a had been as to whether further actions were planned. The workers had seen the Trojan personnel taking measurements and evaluating the smoke problem, but were neither provided any direct information as to the outcome of the initial assessment i
nor were they given an opportunity to disagree with the final PGE assessment, it was not until the problem was encountered with smoke that caused the crane operator to seek medical attention that the contract workers felt Trojan really understood their complaints and took adequate action. The smoke problem may have been an isolated case. Several comments were made concerning how much emphasis was placed on safety at the Trojan site, even to the point that it sometimes slowed down work progress.
For 1997, only one concern had been received by the Excellence Response Program.
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This concern was being evaluated and did not involve an issue in which a work hazard existed. This issue was more a personnelissue. Though only one concern had been received by Trojan's employee concern program, there was evidence that
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numerous other concerns had been identified by employees and expressed to their first line supervision where resolution was underway or had been completed. For the most part, it appeared that this process was working effectively and resulted in the minimal number of problems being reported to the Excellence Response Program.
7.3 CDnclusion The licensee's employee concern program appeared to be receptive to employees raising safety and quality issues. Comments by employees suggested that PGE management was receptive to emp!oyee identified safety and quality problems.
First line supervision had been the main avenue for resolution of day to day problems. This appeared to have been an effective process. However, some weaknesses were noted in the licensee employee concern program related to feedback provided to the workers concerning final resolution of issues. This was discussed with the licensee.
Exit Meeting The inspector presented the inspection results to members of the licensee management and the Resident inspector for the Oregon Office of Energy at the exit meeting on July 24,1997. In addition, a telephone followup exit was concucted on September 8,1997 to present the final NRC resolution for the notice of violation issued in this report. The licensee acknowledged the findings presented. The licensee did not identify as proprietary any information provided to, or reviewed by, the inspector.
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AllaC11MENI PARTIAL LIST OF PERSONS CONTACTED Litanien A. Bowman, Radiation Protection Supervisor J. Cooper, Emergency Preparedness Engineer M. Janouski, Decomrnissioning Engineer T. Meek, Radiation Protection Manager S. Nichols, Decommissioning Projects Manager D. Nordstrom, Nuclear Oversight Manager R. Pato, Licensing Compliance Manager S. Schneider, Operations Manager C. Trimble, Radiation Protection M. Tursa, Engineer G. Zimmerman, Licensing jitate of Orriaon A. Bless, Hesident inspector, Oregon Office of Energy INSPECTION PROCEDURES USED 46051 Structural Concrete Procedure Review 46053 Structural Concrete Work Observation 60851 Design Contro' of ISFSI Components 60854 Preoperational Testing of an ISFSI 71707 Plant Operations 80721 Radiological Environmental Monitoring 83100 Occupational Exposure During SAFSTOR and DECON 83726 Control of Radioactive Material and Contamination 84101 Radioactive Waste Management 86700 Spent Fuel Pool Activities
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ITEMS OPENED, CLOSED, AND DISCUSSED Quene.d 50 344/97003 01 VIO Failure to Perform Adequate Surveys ClDitd i
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Dinuned None i
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LIST OF ACRONYMS ALARA As low As Reasonably Achievable l
CFR Code of Federal Regulations dpm disintegrations per minute GeLi Germanium Lithium HEPA High Efficiency Particulate Filter ISFSI Independent Spent Fuel Storage Installation mR milliroentgen mrem millirem NRC Nuclear Regulatory Commission PGE Portland General Electric RP radiation protection RWP radiation work permits TLD thermoluminescent dosimeters pCi microcurie