IR 05000029/1998001
| ML20236F350 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 06/24/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236F341 | List: |
| References | |
| 50-029-98-01, 50-029-98-02, 50-29-98-1, 50-29-98-2, NUDOCS 9807020126 | |
| Download: ML20236F350 (23) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION 1 Docket No.
50-29 License No.
DPR-03 Report Nos.
50-29/98-01 50-29/98-02 Licensee:
Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398 Facility Name:
Yankee Nuclear Power Station Location:
Rowe, Massachusetts Dates:
January 1 - March 31,1998 Inspectors:
J. Nick, Radiation Specialist, Region i M. Fairtile, Project Manager, NRR L. Pittiglio, Project Manager, NMSS E. Abelquist, ORISE J. Payne, ORISE
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9807020126 980624 I
PDR ADOCK 05000029 G
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e EXECUTIVE SUMMARY
- Yankee Rowe Facility NRC Inspection Report Nos. 50-29/98-01 and 50-29/98-02 These inspections were conducted to determine whether the decommissioning activities carried out at the Yankee Rowe facility were conducted safel and in accordance with NRC
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requirements.' This report covers a three month period of inspection. Areas reviewed included modifications to the spent fuel pool building, radiological controls, and decommissioning activities. = This inspection period also included independent measurements conducted by the Oak Ridge institute for Science and Education (ORISE) on behalf of the NRC to assess the adequacy of the licensee's final status survey plan. These survey results were received by NRC on June 3,1998.
. In general, there were effective programs for protecting the safety of workers during o
dismantlement and decommissioning activities. There was proper implementation of modifications to the spent fuel pool building and good control of radioactive material and contamination with excellent labeling of radioactive material. Overall, the licensee's final status survey program was consistent with NRC guidance. However, several areas for improvement were noted for the final status survey program, particularly regarding assessment
. of scanning sensitivity. The comparison of the licensee's exposure rate measurements to the ORISE survey measurements indicate a potential bias between the pressurized ionization chamber measurements conducted by ORISE and the compensated Geiger-Mueller
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L measurements performed by the licensee..These areas for improvement and additional
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decommissioning program elements are being assessed during the current inspection (6/98),
which also includes ORISE support for independent measurements.
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Report Details
Summary Of Facility Operations On October 12,1995, the Commission served Memorandum and Order CLl-95-14 to Yankee Atomic Electric Company (YAEC) concerning activities at the Yankee Nuclear Power Station (NPS) in Rowe, Massachusetts. The Order states, in part, that "the NRC's approval of the Yankee NPS Decommissioning Plan cannot be accorded further legal effect, pending an
[ adjudicatory] hearing opportunity," and, that in accordance with the pre-1993 interpretation of the decommissioning regulations, "the Commission expects YAEC not to conduct any further ' major' dismantling or decommissioning activities until final approval of its
[ decommissioning] plan after completion of the hearing process." Subsequently, the Commission issued an Order (CLI-96-9), dated October 18,1996, which granted YAEC's Motion for Summary Disposition in a hearing convened to determine whether the decommissioning plan should be approved. Since YAEC had originally submitted the decommissioning plan before the Commission amended its decommissioning regulations, and the decommissioning plan was approved by the NRC in February 1995, YAEC had been given
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approval to conduct decommissioning activities at the Yankee site per a letter from the NRC (reference correspondence, dated October 28,1996, from Mr. Morton Fairtile to Mr. James Kay).
During the decommissioning process, the licensee removed the major primary components from the plant including the pressurizer, four steam generators, and the reactor vessel. Other decommissioning work performed included asbestos abatement, tank removals, piping and conduit removal, removal of the waste evaporator, ano removal of the secondary side equipment (turbine, electrical generator, other components, and piping). The licensee also completed relocation of the control roorn and modifications to allow operation of the spent fuel pool building without dependency on other buildings (spent fuel pool island concept).
When all decommissioning work is completed, the licensee plans to return the site to a " green field" condition with no structures standing on the site.
Operations 02.1 Facility Tours The inspectors toured radiological controlled areas (RCAs) within the vapor containment.
Workers haid removed almost all of the mechanical and electrical components (duct-work, conduit, cables, and fan units) outside the bioshield, in the vapor containment. Workers had also removed most of the remaining structural, mechanical and electrical components from the steam generator and pressurizer cubicles. The stainless steel liner and metal rings from around the reactor vessel opening had been removed from the shield tank cavity. Workers were removing the activated concrete and metal that was left in the area around the reactor vessel opening. The inspectors observed that high radiation area (HRA) controls were satisfactory. All areas were posted and barricaded as required by NRC regulations and Technical Specifications. The posting and labeling of radioactive material were satisfactory, i
Very good radiological controls were provided by health physics technician coverage for
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jobs / activities in RCAs.
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3 The inspectors toured most of the RCAs outside the vapor containment including the primary auxiliary building (PAB), the service building, the radioactive waste processing (compactor)
building, the potentially contaminated area (PCA) storage building (a storage / staging area for potentially contaminated equipment and materials), and the PCA warehouse attached to the radwaste processing building. The PAB had been almost completely cleaned out.
Decontamination and removal of pipe penetrations and pipe chases were continuing.
Underground /underfloor piping was being dug up and removed. Workers had stopped
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decontamination efforts in some areas of the PAB due to a high water table that made low areas fill with water. All radiation areas (ras) and HRAs were posted and barricaded as required. Locked HRAs were maintained locked with appropriate warning signs.
Housekeeping in contaminated areas was very good, and areas that presented a challenge due to work conditions showed continued improvement since the last period of inspection.
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Contamination contol was evident by the use of " step-off pads", personnel monitoring equipment (friskers), and contaminated area postings at the boundaries. Posting and labeling of radioactive materials continue to be excellent. A significant improvement was noted in the previous inspection period due to an increased effort by the licensee's radiation protection staff in labeling radioactive materials to warn workers regarding the potential radiological hazards. No significant safety or NRC regulatory concerns were noted by the inspectors.
O2.2 Current Activities Most mechanical and structural components were removed from the areas in the vapor containment. Workers continued the chipping and removal of concrete around the reactor l
vessel orifice in the shield tank cavity. Decontamination work was continuing in the PAB on the floors, walls and pipe penetrations. Underground pipe removal was started in the PAB, but was stopped due to high water table levels. Modifications to the spent fuel pool building were started. Work was completed for dismantlement of the secondary side equipment in the turbine building and final site surveys had been started. Site characterization activities
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continued and included sub-floor soil sampling. Asbestos abatement was continuing in the former administrative office spaces.
i The licensee was experiencing delays in their decommissioning schedule due to the identification of PCBs in the paint on many metal surfaces. The metal was being temporarily stored in cargo vans at the site while the licensee worked with the Environmental Protection Agency (EPA) regarding removal and disposal methods for the contaminated paint. The EPA approved the decontamination plan for painted surfaces with hazardous material. Review of the decontamination procedures was also planned by the EPA.
Other work planned for 1998 included completion of the items mentioned above, dives in the
spent fuel pool to remove the fuel upender, and an upgrade to the yard' area crane. The final site survey project was slated to be completed near the end of 1998 or early 1999.
10 CFR 50.59 Safety Evaluation Program
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a.
Inspection Scoce (37001)
The inspectors selected for review, the safety evaluation for the Spent Fuel Pool Building j
modifications.
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Observations and Findinss
- The 10 CFR 50.59 and 10 CFR 50.82 evaluations for tNs modification are included in the licensee's Engineering Design Change Request (EPCR) No.96-304. This planned change is -
scheduled to start in Spring 1998 and consists of changes to the building roof and north wall.'
A roof hatch and north wall door have to be enlarged to accommodate a NAC Fuel Transfer Cask in the event the licensee elects to construct an on-site dry cask facility for the temporary -
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storage of spent nuclear fuel. The roof surface is also being replaced. Other modifications to L
. the building are required in order to eliminate obstructions to cask plac6 ment in the pool. The Spent Fuel Pool liner is unaffected by these modifications. ~ The inspectors conducted a walk down of the building in order to examine the structure, hatch, north wall door and the pool
- area in general. The material condition of the building was excellent, no combustible materials
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in the Control Room and determined that all appropriate parameters are being monitored per the.YNPS Defueled Technical Specifications (DTS).
The licensee included, as a part of the EDCR, safety, environmental and ' decommissioning cost
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analyses using the requirements of both.10 CFR 50.59 and 50.82. Subsections 03.1, O3.2, and O3.3, below,'contain the 10 CFR 50.59 information and Subsection 03.4 contains the 50.821nformation.
03.1 Previoualv Analvred Events
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< Section 400 of the YNPS Final Safety Analysis Report (FSAR) assesses the he'alth and safety l j
impacts of decommissioning. The SFP fuel handling accident analysis is provided in Section.
Y E408 of the FSAR. The EPA Protective Action Guides were used as an upper radioactive
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release limit at the site boundary for this accident and all others in the FSAR. ~ The.FSAR also
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assesses a wide spectrum of other accidents including non-radiological events. Loss of spent i
fuel pool cooling, fises, explosions, external events and interactions between decommissioning
" activities and spent fuel were among those analyzed. None of the SFP building modifications L_
- will affect the SFP liner integrity. Based on this, the licensee concluded that none of the
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! above accidents or events would be made worse by the performance of these modifications.
I-The NRC inspectors, based on an independent review, conclude that the consequences of all previously analyzed accidents remain unchanged and that potential radioactive releases *at the site boundary are within the limits of the EPA Protective Action Guides.
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03.2 Accidents of a Tvoe Not Previous!v Analyzed if the licensee, in the future, decides on dry cask storage, the 75-ton capacity yard crane and all lifting devices to be used to handle a loaded cask, will be designed to conform to NUREG-0612, therefore, a heavy load drop need not be considered as the crane and lift components will be single failure proof. The DTS prohibit handling loads in excess of 900 pounds, approximate weight of a single fuel assembly, over spent fuel; however. the drop of a single fuel assembly is an analyzed event and need not be reconsidered. The licensee is constructing a temporary floor to prevent any dropped tools or similar materials from accidentally falling into the pool.
Based on the proposed redesign of the crane and lifting devices to NUREG-0612 standards and the DTS load restrictions, the inspectors conclude that there is no unreviewed accident that needs to be considered.
O3.3 Mamin of Safety as Defined in DTS Bases is Not Reduced The licensee, as required by 10 CFR 50.59, reviewed all the DTS bases related to the SFP, namely, water level, water temperature, loads over the pool, radiation monitoring over pool area and design features to prevent inadvertent pool drainage. the licensee concluded that the margin of safety as defined in the DTS bases will not be reduced by implementation of this EDCR. The inspectors through an independent review and based on the walk down of the pool building find that there is no significant reduction in any margin of safety regarding spent fuel pool operation.
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03.4 Requirements of 10 CFR 50.82(a)(6)
This portion of the regulation prohibits performance of any decommissioning activities that:
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Foreclose release of the site for possible unrestricted use; The purpose of decommissioning is to prepare a site for possible unrestricted use and part of this process is the removal of radioactive materials. The proposed modifications to the Spent Fuel Pool Building are a part of this process.
l The NRC staff has concluded in its Order Approving the Decommissioning Plan
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l and Authorizing Decommissioning of the YNPS of February 14,1995 and as
affirmed by letter dated October 28,1996 from NRC to the licensee, that this l
work is a proper part of the decommissioning of the site; therefore, this modification does not foreclose release of the site for possible unrestricted use.
ii)
Result in significant environmental impacts not previously reviewed; The environmentalimpacts of the decommissioning of the YNPS site were
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addressed in the NRC Environmental Assessment (EA) that was a part of the aforementioned February 14,1995 Order. This EA concluded that YNPS decommissioning was bounded by the NRC Final Generic EnvironmentalImpact Statement on Decommissioning of Nuclear Facilities, NUREG-0586, issued in
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August 1988. Based on the inspector's walk down of the building; review of the EDCR, DTS and FSAR; and Order of February 14, the inspectors conclude
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that the proposed modification will not result in any unreviewed environmental impacts.
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Result in there no longer being reasonable assurance that adequate funds will be available for decommissioning.
The licensee properly included the cost of this modification in cost estimates
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that were approved in the NRC Order of February 14,1995. In addition, there has been no cost overrun in the decommissioning program, to date. Based on
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the above, the inspectors conclude that there is reasonable assurance that l-adequate funds will remain to be available for future decommissioning activities.
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Conclusions Based on this inspection,' which include' visual observation of the Spent Fuel Pool Building d
and Control Room; review of EDCR 96-304, the Decommissioning Order, the FSAR, the DTS
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l and interviews with senior licensee staff, the inspectors conclude that the requirements of 10
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CFR 50.59 and 50.82 are properly implemented in the modification of the Spent Fuel Pool Building.
l R1 Plant Support - Radiological Protection and Chemistry (RP&C) Controls R1.1. Bulk Material Release From YNPS Site a.
insoection Scope (83100f This inspection was performed in order to determine if any bulk licensed radioactive material had been improperly removed from the YNPS site during the period of time from start of reactor operation in 1961 to the present (March 1998).
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Observations and Findinas
,Over a period of five months, ending in early March 1998, the licensee performed a detailed assessment of the YNPS bulk material release program for the 37-year period of interest. The assessment was composed of the following initiatives: (1) Review of history of bulk material releases generated by site excavations, (2) Review of release records, applicable procedures governing material releases, Contamination Summary Reports and Survey Area History Files; and (3) sending questionnaires and conducting follow-up interviews with cognizant current and former plant staff and in addition, following this same procedure with the excavation contractor who provided plant services from _1967 to the present time.
Most of the materials resulting from excavations was disposed of on-site in an area defined as the Southeast Construction' Fill Area. NRC has included this area in its final site release
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A review by the inspectors of the licensee's records indicates that in 1990 and 1993 some soil and asphalt materials from excavations were disposed of off site as low level radioactive waste. 'However, these materials were shipped under NRC regulations to an appropriate low level waste disposal site. Over the 37 years of interest, some quantities of rock, stone, resin, sod and charcoal materials were also released off site; these materials were below the release limits apecified in the licensee's procedures and were therefore properly not classified as " low levei waste." The inspectors reviewed these release procedures going back to the start of operations. The procedures have been continuously updated to conform to changes in NRC regulations and other NRC guidance. As an example, the licensee conservatively incorporated the survey guidance of NRC I & E Circular 81-07 into the YNPS plant material release procedures when the Circular was issued. The surveys listed in the procedures were required to be performed by qualified radiation protection personnel.
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Conclusions Based on the results of this inspection, the inspectors conclude that the removal of bulk radioactive materials from the site, in the time period from start-up through March 1998 was properly performed, using the appropriate plant procedures in force at the time.
R1.2 External Exoosure Control a.
Insoection Scope (83100)
The inspectors reviewed the controls for external radiation exposure through observation of work activities, tours of the facility, interviews with personnel, and a review of licensee documents, b.
Observations and Findinas The inspectors observed the various work activities throughout the facility during the period of this inspection. Personnel in the RCA were observed wearing their assigned thermoluminescent dosimeter (TLD) and pocket ion chamber (PIC) dosimeter. The dose totals for each individual were tallied on each workday and reports were available for review by
. personnel. Plant management periodically reviewed the status of workers in the respective departments.
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Controls for radiation areas and high radiation areas were appropriate throughout the facility.
The inspectors noted that the number of high radiation areas was decreasing, and only seven high radiation areas were currently identified with two being controlled through locked doors / gates (excluding the underwater dose rates from the spent fuel). Radiation work permits (RWPs) and a computerized access control system were also used to control workers'
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radiation exposure. The inspector reviewed selected RWPs written for various work activities
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and concluded that they contained appropriate requirements including administrative dose j
limits, protective clothing, and special monitoring or dosimetry.
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Conclusions
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Controls for extemal radiation were very good including controls used during the various work
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R1.3 Internal Exoosure Control l
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inspection Scooe (83100)
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The inspectors reviewed the controls for internal radiation exposure through observation of work activities, tours of the facility, interviews with personnel and a review of licensee documents, b.
Observations and Findinos
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The inspectors observed operating air sampling equipment in the various areas of the facility
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. during work activities. The equipment was positioned to provide representative sampling of the breathing air in areas occupied by workers. In addition, air handling and filtration equipment was used in areas with potential airborne radioactivity.
The inspectors reviewed the results from internal dose assignments and determined that the.
dose assigned through air sampling and bioassay were very small when compared to the total dose assignment.
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Conclusions i
The licensee had provided good controls for internal radiation exposure including air samphng and bioassay for dose assessment. No violations of NRC regulations and no safety concerns were noted.
R1.4 Control of Radioactive Materials and Contamination Survevs and Monitorino
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Inspection Scope (83100)
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The inspectors reviewed the controls for radioactive materials and contamination, surveys and monitoring through observation of work activities, tours of the facility, interviews with personnel and a review of licensee documents, b.
Observations and Findinog The inspectors verified that there was an adequate supply of radiation survey and monitoring
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equipment. All equipment checked by the inspector was operable and within the current calibration period. Portal monitors and frisking instruments were located throughout the facility for use by workers as they left radioactive materials areas or contaminated areas.
Current radiological surveys of various work locations were reviewed by the inspector. The p
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surveys contained detailed information regarding current dose rates and hazards in the work areas. Surveys were posted at the main control point for the RCA and at the vapor containment control point. Appropriate licensee management personnel had reviewed the radiological surveys.
Radiological housekeeping was good throughout the plant with appropriate controls established to minimize the spread of contamination. As noted in section 02.1 of this Report, areas that presented a challenge to the licensee's staff due to changing conditions and
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ongoing work were kept in a very good condition. These areas (mainly in the lower PAB)
chowed significant improvement in maintenance of the areas in accordance with the licensee management's expectations and to effectively control radioactive contamination. Posting of radioactive material areas and labeling of radioactive materials was excellent. Some inconsistency in container labeling was brought to the attention of licensee representatives during the previous inspection, and the inspectors verified during the current inspection that the labeling had been corrected. Continued good performance was noted by the inspectors as a result of major efforts by the licensee's radiation protection staff.
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Conclusions The licensee provided very good controls for radioactive materials and contamination, surveys and monitoring during decommissioning work activities. No violations or significant safety concerns were identified.
R2 Final Site Survey Plan i
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Insoection Scooe (83801)
The Environrnental Survey and Site Assessment Program (ESSAP) of the Oak Ridge Institute for Science and Education (ORISE) performed decommissioning inspection activities at the YNPS during the period March 9 to 12,1998. Inspection activities included a review of the decommissioning program, with emphasis on the final status survey plan. Survey activities included side-by-side measurements of surface activity and exposure rates.
The NRC inspectors and ORISE staff reviewed YNPS procedures for the final status survey program and several final survey packages.
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Observations and Findinos R2.1 General R2.1.1 Review the past operational radiological surveys that were used to demonstrate l
radiological control of the facility. Are there any records of spills or other releases of radioactive material? If so, do the records adequately document the
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cleanup of these releases of material?
History files are furnished as part of the completed final status survey packages l
being prepared by YNPS staff. Two history files were reviewed during the
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subject inspection Survey Packages OGOO2 and TB015. The basic information
contained in each history file included a survey area description, historical use of the area, and decommissioning /remediation activities. The history file for TB015 (Ground Floor Lube Oil Room in the Turbine Building) included information concerning a contaminating event that occurred in 1967 during rebuilding the main coolant pump. The record indicated _that the maximum removable activity
was 170,000 dpm/100 cm on the operating floor of the Turbine Building. The
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area was subsequently remediated and the record states that recent radiological l
surveys demonstrated contamination levels less than 1,000 dpm/100 cm.
Overall, the history file is complete and provides sufficient justification for the classification of TB015 (which was "affected" in this case).
R2.1.2 Review the results of characterization and other surveys for justification of the classification of the site into affected and unaffected areas.
The initial classification of areas was based on site characterization data, history of radioactive materials use, recommendations by YNPS personnel-knowledgeable of site conditions, and results of operational surveys in support of decommissioning. This classification process is controlled by YNPS administrative procedure AP-8801,"FSS Survey Area Classification and l
Description." This procedure also provides guidance on reclassification of areas as the final status survey progresses. The characterization data used to classify the Northeast Buffers Zone (Survey Package OGOO2) was reviewed and the results of that review follow.
The Northeast Buffers Zone (Survey Package OGOO2) was initially classified as j
unaffected based on no documented use of radioactive materials in this area and -
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that it was not used for site operations or decontamination activities. However, because the characterization data were limited and based on one sample (of nine) exhibiting Co-60 contamination at levels greater than 25% of guideline (sample result was 0.754 pCi/g Co-60), the decision was made on 3/5/97 to reclassify the area as affected. Four months later, this decision was reversed,
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with the reasoning being that the threshold for classification as affected is an individual sample exceeding 75% of guideline, or the average of samples exceeding 25% of guideline. Since this threshold was not met, the survey area was again reclassified as unaffected.
l It appears that this survey area has some potential for contamination in the MARSSIM classification scheme it would likely be a Class 2 area. Currently, this j
L area is classified as unaffected, which means that 30 Soil samples and 10%
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scans will be performed in this survey unit. There was no apparent explanation j
provided for the low, but positive level of Co-60 identified in this area.
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R2.1.3
. Review the specific procedures that are being used to remediate contaminated areas. Consider the potential for incomplete remediation based on these remedial action techniques particularly the potential for the remedial actions to produce areas of localized contamination. What is the procedure for performing and documenting the remedial action support surveys?
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Remedial action support surveys are documented by the Radiation Protection Department at YNPS prior to turnover to the final status survey group. YNPS administrative procedure AP-8810 provides instruction on the turnover of remediated areas to the final status survey program. Specific remediation techniques were not reviewed during this inspection.
R2.2 - Identification of Contaminants and Guidelines
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' R2.2.1 Review the past anblytical results to confirm the nature of the contaminants at the site including the relative ratio of contaminants.
l YNPS used composite samples from the Waste Disposal Building, Primary
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Auxiliary Building and the Vapor Container to determine the nature and relative
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- ratio of contaminants. The contaminants included fission products, activation L
products and transuranic radionuclides (from memo RP 97-39,6/24/97). The predominant radionuclides were Ni-63, Fe-55, Co-60 and Cs-137. YNPS has used these relative ratios 'and the respective guidelines for each radionuclides to develop site-specific guideline values.
R2.2.2 Review the guidelines that the licensee is using for indoor and outdoor areas and.
verify that an appropriate reference has been cited including' surface activity and volumetric guidelines.
For surface activity guidelines, YNPS is referencing the NRC Regulatory Guide 1.86 and the modified surface activity criteria for H-3 and Fe-55 (both hard-to-detect nuclides) approved by the NRC. YNPS is also planning to include Ni-63 in i
the modified surface activity criteria previously approved for only H-3 and Fe-55.
' It is anticipated that YNPS will prepare a technical basis document for this request.
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During the review of YNPS determination of site guideline values, it was noticed that Pu-241 was given a guideline value of 5,000 dpm/100 cm.. This is in
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L apparent conflict with Regulatory Guide 1.86 which assigns a surface activity
limit of 100 dpm/100 cm for transuranic radionuclides.
Soil guidelines were not reviewed during this inspection.
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L R2.2.3 Evaluate how the stated guidelines are being implemented e.g., use of surrogate measurements, presence of multiple contaminants, averaging conditions, and l.
hot spots.
V For surface activity guidelines, YNPS has calculated site-specific guidelines based on the results of the three composite samples discussed in Section R2.2.1. The site-specific surface activity guidelines for the Waste Disposal Building, Primary Auxiliary Building and the Vapor Container were 3,600,3,200
and 4,300 dpm/100 cm, respectively. These guidelines are taken as average values of surface activity, the maximum limit is three times these average limits.
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R2.3 Final Status Survey Procedures and Instrumentation R2.3.1 Review situations where an area's classification was changed based on accumulated survey data from scoping and characterization surveys. Review documentation to determine whether reclassifications were clearly documented.
The classification for the Northeas' Buffers Zone (Survey Package OGOO2) was initially classified as unaffected, than cffected, and is currently an unaffected area once again. The documentation for these reclassifications was reviewed and is described above in section H2.1.2.
R2.3.2 Selection of Background Reference Areas a.
Evaluate the reference area selected by the licensee for background measurements (sampling area should have similar physical, chemical, geological, and radiological characteristics as the site areas being evaluated).
YNPS performed a background soil study at off-site locations to obtain appropriate background reference areas for Cs-137. Four areas were selected and sampled, resulting in a total of approximately 80 soil samples. For disturbed soil areas, YNPS stated that it will apply a zero Cs-137 background (it is expected that this will apply to soil beneath the asphalt in affected areas).
Background surface activity measurements were collected from the Screenwell House. YNPS uses background levels from both generic and non-generic i
materials. Generic materials are those materials that exhibit the same background level as their environment, and include steel, glass, lumber, carpet, i
etc.
b.
Verify'that the background reference area has not been affected by site operations, including waste management practices and process effluents.
The background reference areas selected do not appear to have been affected l
by site operations.
c.
Determine whether the licensee has performed sufficient background sample analyses to adequately assess the true background level and its variability.
i Particularly for background soil concentrations, YNPS has collected sufficient samples to evaluate the true background levels and their variability. The supporting documentation for surf ace material background levels was not
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reviewed during this inspection.
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R2.3.3 -
Selection of Survey Instrumentation
~ R2.3.3.1
. Land Areas a.
' Review the basis for the selection of instruments [e.g., based on potential contaminants and their associated radiations, types of media (soil, sludge, etc.),
. to be ~ evaluated, and detection sensitivities].' Typically,' Nal scintillation detectors are used for land area surveys.
' Appropriate instrumentation is being used for the final status survey for outdoor i
land areas, b.
Review documentation pertaining to instrumentation sensitivity, particularly.
licensee statements to the effect that instrumentation will be sufficient to detect radiological contamination.: The detection sensitivity should be below the appropriate guideline values.
.
YNPS has not provided estimates of their instruments' detection sensitivity for
)
outdoor scans of land areas. Discussed with the licensee that nominal values i
for scan MDC are now provided in NUREG-1507 and MARSSIM.
c.
- Evaluate the instrument scan sensitiv.ty for scan surveys of land ' areas. Check the scan sensitivity in terms of the soil guideline.
See above response.
Evaluate methods and procedures for exposure rate measurements.
. YNPS uses a compensated GM to perform exposure rate measurements.- A complete evaluation, including derivation of their correction coefficient, was not performed during this inspection, i
i e.
Evaluate the procedures for instrument calibration (use of appropriate
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radionuclides calibration sources, source geometry,'and appropriate consideration i
of environmental conditions).
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Not applicable.
f.
Review the operational check-out of survey instrumentation. Evaluate frequency of operational checks (both to calibration source and background) and if
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instrument response fell within predetermined acceptance criteria.
L YNPS' performs operational source response checks and background response
. checks of their instrumentation consistent with industry standards. These E
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response check procedures are provided in DP-8832, "FSS Control and L Accountability of. Portable Survey. instruments." However, the frequency of these response c"ecks is unclear. It is recommended that these response p
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' checks be performed prior to use and the completion of use, each day that the instrumenu 3rs used.
l R2.3.3.2 Building Surfaces
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a.
Review the basis for the selection of instruments (e.g., based on potential contaminants and their associated radiations, surface types to be evaluated, and detection sensitivities). Typically, GM or gas proportional detectors are used for building surface contamination surveys.
Appropriate instrumentation is being used for the final status survey for building surfaces. _YNPS plans to employ both GM or gas proportional detectors for the.
FSS.
b.
Review documentation pertaining to instrumentation sensitivity, particularly
,
licenses statements to the effect that instrumentation will be sufficient to detect
,h radiological contamination. The detection sensitivity should be below the l
appropriate guideline values.
YNPS provided detection sensitivities in the Final Status Survey Plan. All
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-instruments used for direct measurements of surface activity are sufficiently l
_below guideline values.
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Evaluate the instrument scan sensitivity for interior scan surveys.
c.
YNPS provided scan MDCs on Table 2 of the FSS Plan for GM or gas proportional detectors, including the floor monitor. Instruments used for surface
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scans are generally below the site-specific guideline values.
d.1 Evaluate the procedures for instrument calibration (use of appropriate
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radionuclides calibration sources, source geometry, and appropriate consideration l-of surface and environmental conditions).
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YNPS calibrates their surface activity measurement instrumentation using a Co-L 60 standard, at a distance of 0.5 inches from the source. This radionuclides -
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source is appropriate based ort the nature of the contamination present at the
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site (based on composite samples and process knowledge). Specifically, the Co-
2 l;
60 source has an area of 100 cm and has a 1.7 mg/cm cover over the source,
.which helps to account for surface and environmental conditions.
e..
Review the licensee's MDC equation for direct measurements on building -
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surfaces.
YNPS uses an appropriate MDC equation for direct measurements of surface activity (reference NUREG-1507).
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _
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- f.
Review the operational check-out of survey instrumentation [ evaluate frequency l
of operational checks (both to calibration source and background) and if E
instrument response fell within predetermined acceptance criteria).
YNPS performs operational source response checks and background response
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checks of their instrumentation consistent with industry standards. These i
. response check procedures are provided in DP-8832, "FSS Control and Accountability of Portable Survey instruments." However, the frequency of these response checks is unclear.._ lt is recommended that these response -
checks be performed prior to use and the completion of use, each day that the instruments are used.
l R2.3.4.
Determine the use of investigation levels for measurement results and if the l
h licensee performed appropriate follow-up actions. For example, soil samples should be collected if the Nal scintillation detector readings exceed a specified investigation level.
Table 5.1 in the FSS Plan provides appropriate investigation levels for both surface activity and soil contamination.
R2.3.51 Survey Procedures.
a.
Review the procedure for performing surface activity measurements and surface scans on building surfaces.
Indoor survey procedures are consistent with guidance in NUREG/CR-5849.
' b.
Review the procedures for performing soil samplina and outdoor surface
scanning. ' Identify sampling methods and equipment, implementation requirements, decontamination procedures for sampling equipment, and sample i.-
chain-of-custody procedures.
Outdoor survey activities were not reviewed during this inspection.
R2.3.6.
When using data from scoping or characterization surveys as final status survey data in an area, what procedures are in place to ensure that the radiolog cal
conditions have not changedT.
YNPS procedure AP-8810, AFSS Survey Area Turnover and Control describes l-the approach used to assure that survey areas are not decontaminated by i
personnel or decommissioning activities in nearby areas. This procedure l-
- provides for routine surveys of areas where FSS have been completed and lists l_
access control measures following completion of the FSS.
R2.4 Analytical Procedures For Soil Samoles
. R2.4.1 Review the licensee's laboratory analytical procedures for radiological analyses.
Specifically:
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Evaluate the laboratory's sample preparation techniques geometries used for gamma spectrometry on soil samples, etc.
b.
Review the protocol the lab uses to interpret the gamma spectrometry results, particularly the radionuclides photopeaks used to identify various contaminants.
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c.
Review the laboratory QA/QC procedures, including duplicates, blanks, and matriy spikes. Determine the frequency of analysis for each of the QC checks.
Determine whether the lab participates in some sort of cross-check or performance evaluation program, such as that offered by EML and EPA.
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This inspection area will be covered during a future inspection at YNPS.
J R2.5 Instrument Comparison Activities
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Review of surface activity measurements using gas proportional and GM detectors, a.
The specific material background radiation levels used to correct the gross counts will be determined for a number of surface types in unaffected areas at the site. These areas will be considered to be representative of the areas included in the final status survey. The surfaces willinclude various types of concrete, concrete block walls, brick surfaces, sheetrock walls, tiled floors, steel 1-beams, et cetera. The method used by the licensee to determine background levels will be evaluated, and the results obtained by ORISE and the licensee compared.
ORISE obtained background ' measurements on poured concrete and unpainted concrete block from the Screenwell House using a gas proportional detector.
YNPS performed similar measurements on these materials using their gas proportional detector. ORISE obtained average count rates of 389 and 350 cpm
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l for the poured concrete and unpainted block, respectively, while YNPS r
measured average count rates of 371 and 331 cpm for the poured concrete and unpainted block.
b.
. Detector calibration will be evaluated, particularly the calibration source radionuclides, geometry of the source and source-to-detector spacing, window density thickness, and any other factors considered to have a measurable effect on the detector efficiency.
YPNS calibrates its detectors in a jig set to maintain a 0.5 inch spacing. This same spacing is used during field measurements of surface activity. YNPS uses
2
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. a shielded (with 1.7 mg/cm ) Co-60 source with an area of 100 cm. YNPS
l uses gas proportional detectors with a 0.96 mg/cm window density thickness
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and the nominal detection efficiency for Co-60is 14%. All of these factors are appropriate for the measurement of surface activity at YNPS.
c, The minimum detectable concentration (MDC) equations will be compared and discussed. Discussion topics willinclude background and sample count times.
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Both ORISE and YNPS use 1-minute count times end the resulting MDC calculations are comparable, i-d.
The interpretation of the raw survey data will be evaluated. Specific discussion items willinclude subtraction of the appropriate backgrounds, probe area corrections, and averaging areas for measurement data (e.g., the manner in which surface activity results between one and three times the guideline will be treated).
- YNPS uses a probe area correction of 1 (100 cm probe area). The averaging area for direct measurements is based on the guidance in NUREG/CR-5849 and
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is 1 m'.
e.
Following consideration of items a to d above, side-by-side field measurements will be performed by ORISE and the licensee, within affected areas, to compare actual surface activity levels. Both gross and net count rates will be compared, i
to determine the effect of background on the results.
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ORISE and YNPS performed side-by-side measurements in the Hot Machine Shop 20 locations. The gross counts compared favorably. Eleven of the measurements were performed on concrete. Background values determined in the Screenwell House were applied to these values and surface activity levels were calculated (Table 1). In order to calculate surface activity levels, ORISE j
determined its efficiency on YNPS' Co-60 source, at a distance of 0.5 inches-resulting in a 12.7% efficiency. However, ORISE measurements were
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performed at contact, so an estimate of the ORISE efficiency at contact with YNPS Co-60 source was performed. Data in NUREG-1507 on reduction of efficiency with increasing distance from the source for Tc-99 was used. The calculated efficiency was 19.3%. Table 1 shows the results of surface activity calculations for ORISE (both using (NPS source and standard ORISE efficiency
determination) and YNPS.
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R2.5.2 :
Exposure rate measurements using a PIC.
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. Background exposure rate levels will be de'termined for both indoor and outdoor
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F
~ areas within unaffected areas.
These areas - will be considered ' to be t
.~ representative'of the areas included in the final status survey. Data will be l,
reported in R/h and the results obtained by ORISE and th' licensee compared.
s
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b.
Side-by-side field measurements of exposure rate will be performed by ORISE (-
and the licensee, within indoor and outdoor affected areas, to determine actual exposure rates.
f.
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p ORISE and the licensee performed indoor exposure rate measurements to permit B
. data comparison. ' Exposure rates were measured at one meter above the surface, with ORISE using a pressurized ionization chamber (PIC) and YNPS
,
using a compensated GM detector..
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L A background. exposure rate measurement was performed on the upper level of the unaffected Screenwell House.' ORISE measured a background exposure rate of 8.4pR/h,.YNPS measured a background exposure rate of 9.3 pR/h.
Field exposum rates were measured at two locations within the affected Hot Machine Shop, one in the North Decon Room and one in the Maintenance Shop area. The ORISE measured exposure rates for the North Decon Room and the, L
. Maintenance Shop area were 134 and 27.1 pR/h, respectively. YNPS measured
>
p 102.7 and 22.7pR/h, respectively. ORISE appears to be biased high relative l
to YNPS.
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. Outdoor exposure rates in affected and unaffected areas will be measured during a future inspection at YNPS.
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- R2.5.3.
Miscellaneous items
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a'
Smears for removable surface contamination will be discussed, particularly for the assessment of: H-3 and Fe-55.
Topics for discussion will include L
methodology' for ' sample (smear) collection, e.g., moistened smears and y
laboratory analysis procedures.
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l YNPS has no plans currently for performing smears for the assessment of H-3
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and Fe-55..The purpose for performing these assessments is to assure that the radionuclides mixture determined by three composite samples is appropriate in all areas, particularly plant systems. YNPS has stated in the FSS Plan that as
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further activities are performed, the radionuclides mix will be' evaluated.
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.b.
General discussion of instruments and procedures used for scanning (topics will include scanning sensitivity and background levels).
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Discussions with the licensee pertaining to scan MDC. Refer to section 3.3.1 (b).
c.
Instrumentation. and procedures for performing alpha surface activity measurements. Discuss situation where alpha and beta surface activity may be present and the use of alpha blockers (3.8 mg/cm Mylar).
1 YNPS has no plans for performing alpha surface activity measurements. The purpose for performing alpha measurements is to assure that the radionuclides mixture determined by three composite samples is appropriate in all areas, particularly any areas that may have a higher potential for alpha contamination.
YNPS has stated in the FSS Plan that as further activities are performed, the radionuclides mix will be evaluated.
c.
1 Conclusions Based on the subject decommissioning inspection, the following concerns were identified:
e YNPS is planning to include Ni-63 in the modified surface activity criteria previously
[
approved for only H-3 and Fe-55. YNPS has subsequently prepared a technical basis document for this request.
el Side-by-side comparisons of exposure rate measurements indicate a potential bias between the Plc measurements performed by ORISE and compensated GM
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' measurements performed by YNPS. Additional discussion and measurements are
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recommended to resolve this apparent discrepancy.
e.
The Northeast Buffers Zone (Survey Package OGOO2) was initially classified as unaffected, then reclassified as affected, and currently this survey area is again classified as' unaffected.
While justifications are;provided for each of.these classifications, there is no apparent explanation provided for the low, but positive level of Co-60 identified in this area, t
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> Although the final survey packages were not finalized, the following items were identified as
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' a result of the review:
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e
. Some packages contained other information in the survey area historical file that did j
not provide accurate historical information to support the classification of the area.
i e
Changes to the survey package were not always signed and dated.
e
.The licensee procedures provided a description of the content of the survey package, but did not list the contents that are required (or optional) for.each survey package.
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By defining the content 'of the survey package in the procedure, the licensee could
'
' ensure that the packages contained the same types of information.
I o
The NRC and the licensee's staff discussed the benefit for providing a summary or release record to make it easier to review the contents of a survey package.
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Overall, the YNPS final status survey program is consistent with the recommendations of NUREG/CR-5849 and pertinent industry standards. ORISE made several recommendations-for improving the final status survey program, particularly regarding assessment of scan sensitivity. It is anticipated that additional decommissioning program elements will be assessed during the next inspection.
R8 Miscellaneous issues R8.1 Review of Uodated Final Safety Analysis Report (UFSAR) Commitments l
A recent discovery of a licensee operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the UFSAR description.
While performing the inspections ' discussed in this report, the inspector reviewed the applicable portions of the UFSAR that related to the areas inspected. The inspector verified that the UFSAR wording was consistent with the observed plant practices, procedures, and/or parameters.
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X1 Exit Meeting Summary The inspectors met with the licensee representatives denoted below at the conclusion of the on-site inspection on January 15,1998 and March 12,1998. The inspectors summarized the i
purpose, scope, and findings of the inspection. The licensee representatives acknowledged
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the inspection findings.
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1 PARTIAL LIST OF PERSONS CONTACTED
- 'G. Babineau, Radiation Protection and Chemistry Manager
- 'W. Blackadar, Radiation Protection Engineer H. Breite, Lead Engineer
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M. Clary, Design Engineer j
K. Corbett, Quality Assurance
- W. Cox, Radiation Protection Engineer
E. Cummings, Lead Engineer j
C. Ellis, Radiological Engineer i
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- S. Garvie, Security Supervisor
- J. Grant, Decommissioning Manager
- R. Greenfeld, Radiation Protection Engineer /ALARA Program
- R. Grippardi, Quality Assurance Supervisor E. Heath, Final Status Manager
K. Heider, Site Manager
S. Litchfield, Health and Safety Supervisor
- 'C. Melin, Construction Manager S. Mullet, Radiation Protection Technician
S. Roberts, Radiological Engineer i
- A. Trudeau, Quality Services Group Senior Engineer M. Vandale, Radiation Protection Senior Engineer
- 'F. Williams, Plant Superintendent
- G. Wood, Instrumentation Specialist
Denotes those individuals participating in the exit briefing held on January 15,1998
Denotes those individuals participating in the exit briefing held on March 12,1998 INSPECTION PROCEDURES USED IP 83100: Occupational Radiation Exposure During Decommissioning IP 86750: Solid Radwaste Management and Transportation IP 84750: Effluent and Environmental Monitoring IP 81700: Physical Security Program for Power Reactors
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I ITEMS OPENED, CLOSED, AND DISCUSSED
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